Transmission electron microscopy investigation of neutron irradiated Si and ZrN coated UMo particles prepared using FIB

NASA Astrophysics Data System (ADS)

Van Renterghem, W.; Miller, B. D.; Leenaers, A.; Van den Berghe, S.; Gan, J.; Madden, J. W.; Keiser, D. D.

2018-01-01

Two fuel plates, containing Si and ZrN coated U-Mo fuel particles dispersed in an Al matrix, were irradiated in the BR2 reactor of SCK•CEN to a burn-up of ∼70% 235U. Five samples were prepared by INL using focused ion beam milling and transported to SCK•CEN for transmission electron microscopy (TEM) investigation. Two samples were taken from the Si coated U-Mo fuel particles at a burn-up of ∼42% and ∼66% 235U and three samples from the ZrN coated U-Mo at a burn-up of ∼42%, ∼52% and ∼66% 235U. The evolution of the coating, fuel structure, fission products and the formation of interaction layers are discussed. Both coatings appear to be an effective barrier against fuel matrix interaction and only on the samples having received the highest burn-up and power, the formation of an interaction between Al and U(Mo) can be observed on those locations where breaches in the coatings were formed during plate fabrication.

Low-temperature irradiation behavior of uranium-molybdenum alloy dispersion fuel

NASA Astrophysics Data System (ADS)

Meyer, M. K.; Hofman, G. L.; Hayes, S. L.; Clark, C. R.; Wiencek, T. C.; Snelgrove, J. L.; Strain, R. V.; Kim, K.-H.

2002-08-01

Irradiation tests have been conducted to evaluate the performance of a series of high-density uranium-molybdenum (U-Mo) alloy, aluminum matrix dispersion fuels. Fuel plates incorporating alloys with molybdenum content in the range of 4-10 wt% were tested. Two irradiation test vehicles were used to irradiate low-enrichment fuels to approximately 40 and 70 at.% 235U burnup in the advanced test reactor at fuel temperatures of approximately 65 °C. The fuel particles used to fabricate dispersion specimens for most of the test were produced by generating filings from a cast rod. In general, fuels with molybdenum contents of 6 wt% or more showed stable in-reactor fission gas behavior, exhibiting a distribution of small, stable gas bubbles. Fuel particle swelling was moderate and decreased with increasing alloy content. Fuel particles with a molybdenum content of 4 wt% performed poorly, exhibiting extensive fuel-matrix interaction and the growth of relatively large fission gas bubbles. Fuel particles with 4 or 6 wt% molybdenum reacted more rapidly with the aluminum matrix than those with higher-alloy content. Fuel particles produced by an atomization process were also included in the test to determine the effect of fuel particle morphology and microstructure on fuel performance for the U-10Mo composition. Both of the U-10Mo fuel particle types exhibited good irradiation performance, but showed visible differences in fission gas bubble nucleation and growth behavior.

Modeling a failure criterion for U–Mo/Al dispersion fuel

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

Oh, Jae-Yong; Kim, Yeon Soo; Tahk, Young-Wook

2016-05-01

The breakaway swelling in U-Mo/Al dispersion fuel is known to be caused by large pore formation enhanced by interaction layer (IL) growth between fuel particles and Al matrix. In this study, a critical IL thickness was defined as a criterion for the formation of a large pore in U-Mo/Al dispersion fuel. Specifically, the critical IL thickness is given when two neighboring fuel particles come into contact with each other in the developed IL. The model was verified using the irradiation data from the RERTR tests and KOMO- 4 test. The model application to full-sized sample irradiations such as IRISs, FUTURE,more » E-FUTURE, and AFIP-1 tests resulted in conservative predictions. The parametric study revealed that the fuel particle size and the homogeneity of the fuel particle distribution are influential for fuel performance.« less

A cellular automaton method to simulate the microstructure and evolution of low-enriched uranium (LEU) U–Mo/Al dispersion type fuel plates

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

Drera, Saleem S.; Hofman, Gerard L.; Kee, Robert J.

Low-enriched uranium (LEU) fuel plates for high power materials test reactors (MTR) are composed of nominally spherical uranium-molybdenum (U-Mo) particles within an aluminum matrix. Fresh U-Mo particles typically range between 10 and 100 mu m in diameter, with particle volume fractions up to 50%. As the fuel ages, reaction-diffusion processes cause the formation and growth of interaction layers that surround the fuel particles. The growth rate depends upon the temperature and radiation environment. The cellular automaton algorithm described in this paper can synthesize realistic random fuel-particle structures and simulate the growth of the intermetallic interaction layers. Examples in the presentmore » paper pack approximately 1000 particles into three-dimensional rectangular fuel structures that are approximately 1 mm on each side. The computational approach is designed to yield synthetic microstructures consistent with images from actual fuel plates and is validated by comparison with empirical data on actual fuel plates. (C) 2014 Elsevier B.V. All rights reserved.« less

Fuel swelling and interaction layer formation in the SELENIUM Si and ZrN coated U(Mo) dispersion fuel plates irradiated at high power in BR2

NASA Astrophysics Data System (ADS)

Leenaers, A.; Van den Berghe, S.; Koonen, E.; Kuzminov, V.; Detavernier, C.

2015-03-01

In the framework of the SELENIUM project two full size flat fuel plates were produced with respectively Si and ZrN coated U(Mo) particles and irradiated in the BR2 reactor at SCK•CEN. Non-destructive analysis of the plates showed that the fuel swelling profiles of both SELENIUM plates were very similar to each other and none of the plates showed signs of pillowing or excessive swelling at the end of irradiation at the highest power position (local maximum 70% 235U). The microstructural analysis showed that the Si coated fuel has less interaction phase formation at low burn-up but at the highest burn-ups, defects start to develop on the IL-matrix interface. The ZrN coated fuel, shows a virtual absence of reaction between the U(Mo) and the Al, up to high fission densities after which the interaction layer formation starts and defects develop in the matrix near the U(Mo) particles. It was found and is confirmed by the SELENIUM (Surface Engineering of Low ENrIched Uranium-Molybdenum) experiment that there are two phenomena at play that need to be controlled: the formation of an interaction layer and swelling of the fuel. As the interaction layer formation occurs at the U(Mo)-matrix interface, applying a diffusion barrier (coating) at that interface should prevent the interaction between U(Mo) and the matrix. The U(Mo) swelling, observed to proceed at an accelerating rate with respect to fission density accumulation, is governed by linear solid state swelling and fission gas bubble swelling due to recrystallization of the fuel. The examination of the SELENIUM fuel plates clearly show that for the U(Mo) dispersion fuel to be qualified, the swelling rate at high burn-up needs to be reduced.

Effects of irradiation on the microstructure of U-7Mo dispersion fuel with Al-2Si matrix

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Jue, Jan-Fong; Robinson, Adam B.; Medvedev, Pavel; Gan, Jian; Miller, Brandon D.; Wachs, Daniel M.; Moore, Glenn A.; Clark, Curtis R.; Meyer, Mitchell K.; Ross Finlay, M.

2012-06-01

The Reduced Enrichment for Research and Test Reactor (RERTR) program is developing low-enriched uranium U-Mo dispersion fuels for application in research and test reactors around the world. As part of this development, fuel plates have been irradiated in the Advanced Test Reactor and then characterized using optical metallography (OM) and scanning electron microscopy (SEM) to determine the as-irradiated microstructure. To demonstrate the irradiation performance of U-7Mo dispersion fuel plates with 2 wt.% Si added to the matrix, fuel plates were tested to moderate burnups at intermediate fission rates as part of the RERTR-6 experiment. Further testing was performed to higher fission rates as part of the RERTR-7A experiment, and very aggressive testing (high temperature, high fission density, and high fission rate) was performed in the RERTR-9A, RERTR-9B, and AFIP-1 experiments. As-irradiated microstructures were compared to those observed after fabrication to determine the effects of irradiation on the microstructure. Based on comparison of the microstructural characterization results for each irradiated sample, some general conclusions can be drawn about how the microstructure evolves during irradiation: there is growth during irradiation of the fuel/matrix interaction (FMI) layer created during fabrication; Si diffuses from the FMI layer to deeper depths in the U-7Mo particles as the irradiation conditions are made more aggressive; lowering of the Si content in the FMI layer results in an increase in the size of the fission gas bubbles; as the FMI layer grows during irradiation, more Si diffuses from the matrix to the FMI layer/matrix interface; and interlinking of fission gas bubbles in the fuel plate microstructure that may indicate breakaway swelling is not observed.

Effect of stress evolution on microstructural behavior in U-Mo/Al dispersion fuel [Effect of stress on microstructural evolution in U-Mo/Al dispersion fuel

DOE PAGES

Jeong, G. Y.; Kim, Yeon Soo; Jamison, L. M.; ...

2017-02-20

U-Mo/Al dispersion fuel irradiated to high burnup at high power (high fission rate) exhibited microstructural changes such as deformation of the fuel particles, pore growth, and rupture of the Al matrix. The driving force for these microstructural changes was meat swelling caused by a combination of fuel particle swelling and interaction layer growth. Five miniplates with well-recorded fabrication data and irradiation conditions were selected, and their PIE data was analyzed. ABAQUS finite element analysis (FEA) was utilized to simulate the microstructural evolution of the plates. Using the simulation results shear stress, effective stress and hydrostatic stress exerted on both themore » fuel particles and the Al matrix were determined. The effects of fabrication and irradiation variables on stress-induced microstructural evolutions, such as pore growth in the interaction layers and Al matrix rupture, were investigated. The observed microstructural changes were consistent with the calculated stress distribution in the meat.« less

Effect of stress evolution on microstructural behavior in U-Mo/Al dispersion fuel [Effect of stress on microstructural evolution in U-Mo/Al dispersion fuel

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

Jeong, G. Y.; Kim, Yeon Soo; Jamison, L. M.

U-Mo/Al dispersion fuel irradiated to high burnup at high power (high fission rate) exhibited microstructural changes such as deformation of the fuel particles, pore growth, and rupture of the Al matrix. The driving force for these microstructural changes was meat swelling caused by a combination of fuel particle swelling and interaction layer growth. Five miniplates with well-recorded fabrication data and irradiation conditions were selected, and their PIE data was analyzed. ABAQUS finite element analysis (FEA) was utilized to simulate the microstructural evolution of the plates. Using the simulation results shear stress, effective stress and hydrostatic stress exerted on both themore » fuel particles and the Al matrix were determined. The effects of fabrication and irradiation variables on stress-induced microstructural evolutions, such as pore growth in the interaction layers and Al matrix rupture, were investigated. The observed microstructural changes were consistent with the calculated stress distribution in the meat.« less

Irradiated microstructure of U-10Mo monolithic fuel plate at very high fission density

NASA Astrophysics Data System (ADS)

Gan, J.; Miller, B. D.; Keiser, D. D.; Jue, J. F.; Madden, J. W.; Robinson, A. B.; Ozaltun, H.; Moore, G.; Meyer, M. K.

2017-08-01

Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (<20% U-235 enrichment) as a result of its high uranium loading capacity compared to that of U-7Mo dispersion fuel. These fuel plates contain a Zr diffusion barrier between the U-10Mo fuel and Al-6061 cladding that suppresses the interaction between the U-Mo fuel foil and Al alloy cladding that is known to be problematic under irradiation. Different methods have been employed to fabricate monolithic fuel plates, including hot-rolling with no cold-rolling. L1P09T is a hot-rolled fuel plate irradiated to high fission density in the RERTR-9B experiment. This paper discusses the TEM characterization results for this U-10Mo/Zr/Al6061 monolithic fuel plate (∼59% U-235 enrichment) irradiated in Advanced Test Reactor at Idaho National Laboratory with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 °C, respectively. TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (>1 μm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ∼30 at% and ∼7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.

Amorphization of the interaction products in U-Mo/Al dispersion fuel during irradiation

NASA Astrophysics Data System (ADS)

Ryu, Ho Jin; Kim, Yeon Soo; Hofman, G. L.

2009-04-01

The microstructures of the product resulting from interaction between U-Mo fuel particles and the Al matrix in U-Mo/Al dispersion fuel are discussed. We analyzed the available characterization results for the Al matrix dispersion fuels from both the out-of-pile and in-pile tests and examined the difference between these results. The morphology of pores that form in the interaction products during irradiation is similar to the porosity previously observed in irradiation-induced amorphized uranium compounds. The available diffraction studies for the interaction products formed in both the out-of-pile and in-pile tests are analyzed. We have concluded that the interaction products in the U-Mo/Al dispersion fuel are formed as an amorphous state or become amorphous during irradiation, depending on the irradiation conditions.

Irradiated microstructure of U-10Mo monolithic fuel plate at very high fission density

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

Gan, J.; Miller, B. D.; Keiser, D. D.

Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (< 20% U-235 enrichment) as a result of its high uranium loading capacity compared to that of U-7Mo dispersion fuel. These fuel plates contain a Zr diffusion barrier between the U-10Mo fuel and Al-6061 cladding that suppresses the interaction between the U-Mo fuel foil and Al alloy cladding that is known to be problematic under irradiation. This paper discusses the TEM results of the U-10Mo/Zr/Al6061 monolithic fuel plate (Plate ID: L1P09T, ~ 59% U-235 enrichment) irradiated in Advancedmore » Test Reactor at Idaho National Laboratory as part of RERTR-9B irradiation campaign with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 C, respectively. A total of 5 TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (> 1 µm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ~ 30 at% and ~ 7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.« less

Thermal properties of U-7Mo/Al dispersion fuel

NASA Astrophysics Data System (ADS)

Cho, Tae Won; Kim, Yeon Soo; Park, Jong Man; Lee, Kyu Hong; Kim, Sunghwan; Lee, Chong-Tak; Yang, Jae Ho; Oh, Jang Soo; Won, Ju-Jin; Sohn, Dong-Seong

2017-12-01

The thermal diffusivity and heat capacity of U-7Mo/Al and U-7Mo/Al-5Si as functions of U-Mo fuel volume fraction and temperature were measured. The density of the sample was measured at room temperature and estimated using thermal expansion data at elevated temperatures. Using the measured data, the thermal conductivity was obtained as a function of U-Mo volume fraction and temperature. The thermal conductivity of U-7Mo/Al-5Si was found to be lower than that of U-7Mo/Al because of the Si addition to the Al. Due to a lower porosity and reduced interaction between U-Mo and Al in the sample, the thermal conductivity data reported in the present study were higher than those in the literature. The present data were found to be in agreement with the predictions of theoretical models.

A model to predict thermal conductivity of irradiated U-Mo dispersion fuel

NASA Astrophysics Data System (ADS)

Burkes, Douglas E.; Huber, Tanja K.; Casella, Andrew M.

2016-05-01

Numerous global programs are focused on the continued development of existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world's remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. Some of these programs are focused on assisting with the development and qualification of a fuel design that consists of a uranium-molybdenum (U-Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. Thermal conductivity is an important consideration in determining the operational temperature of the fuel and can be influenced by interaction layer formation between the dispersed phase and matrix and upon the concentration of the dispersed phase within the matrix. This paper extends the use of a simple model developed previously to study the influence of interaction layer formation as well as the size and volume fraction of fuel particles dispersed in the matrix, Si additions to the matrix, and Mo concentration in the fuel particles on the effective thermal conductivity of the U-Mo/Al composite during irradiation. The model has been compared to experimental measurements recently conducted on U-Mo/Al dispersion fuels at two different fission densities with acceptable agreement. Observations of the modeled results indicate that formation of an interaction layer and subsequent consumption of the matrix reveals a rather significant effect on effective thermal conductivity. The modeled interaction layer formation and subsequent consumption of the high thermal conductivity matrix was sensitive to the average dispersed fuel particle size, suggesting this parameter as one of the most effective in minimizing thermal conductivity degradation of the composite, while the influence of Si additions to the matrix in the model was highly dependent upon irradiation conditions.

Effects of Irradiation on the Microstructure of U-7Mo Dispersion Fuel with Al-2Si Matrix

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

Dennis D. Keiser, Jr.; Jan-Fong Jue; Adam B. Robinson

2012-06-01

The Reduced Enrichment for Research and Test Reactor program is developing low-enriched uranium U-Mo dispersion fuels for application in research and test reactors around the world. As part of this development, fuel plates have been irradiated in the Advanced Test Reactor and then characterized using optical metallography (OM) and scanning electron microscopy (SEM) to determine the as-irradiated microstructure. To demonstrate the irradiation performance of U-7Mo dispersion fuel plates with 2 wt% Si added to the matrix, fuel plates were tested to medium burnups at intermediate fission rates as part of the RERTR-6 experiment. Further testing was performed to higher fissionmore » rates as part of the RERTR-7A experiment, and very aggressive testing (high temperature, high fission density, high fission rate) was performed in the RERTR-9A, RERTR-9B and AFIP-1 experiments. As-irradiated microstructures were compared to those observed after fabrication to determine the effects of irradiation on the microstructure. Based on comparison of the microstructural characterization results for each irradiated sample, some general conclusions can be drawn about how the microstructure evolves during irradiation: there is growth of the fuel/matrix interaction layer (FMI), which was present in the samples to some degree after fabrication, during irradiation; Si diffuses from the FMI layer to deeper depths in the U-7Mo particles as the irradiation conditions are made more aggressive; lowering of the Si content in the FMI layer results in an increase in the size of the fission gas bubbles; as the FMI layer grows during irradiation more Si diffuses from the matrix to the FMI layer/matrix interface, and interlinking of fission gas bubbles in the fuel plate microstructure that may indicate breakaway swelling is not observed.« less

A model to predict thermal conductivity of irradiated U–Mo dispersion fuel

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

Burkes, Douglas E.; Huber, Tanja K.; Casella, Andrew M.

The Office of Materials Management and Minimization Reactor Conversion Program continues to develop existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world’s remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. The program is focused on assisting with the development and qualification of a fuel design that consists of a uranium-molybdenum (U-Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. Thermal conductivity is an important consideration in determining the operational temperature of the fuel and can be influenced by interaction layermore » formation between the dispersed phase and matrix and upon the concentration of the dispersed phase within the matrix. This paper extends the use of a simple model developed previously to study the influence of interaction layer formation as well as the size and volume fraction of fuel particles dispersed in the matrix, Si additions to the matrix, and Mo concentration in the fuel particles on the effective thermal conductivity of the U-Mo/Al composite during irradiation. The model has been compared to experimental measurements recently conducted on U-Mo/Al dispersion fuels at two different fission densities with acceptable agreement. Observations of the modeled results indicate that formation of an interaction layer and subsequent consumption of the matrix reveals a rather significant effect on effective thermal conductivity. The modeled interaction layer formation and subsequent consumption of the high thermal conductivity matrix was sensitive to the average dispersed fuel particle size, suggesting this parameter as one of the most effective in minimizing thermal conductivity degradation of the composite, while the influence of Si additions to the matrix in the model was highly dependent upon irradiation conditions.« less

Atomistic Simulation of High-Density Uranium Fuels

DOE PAGES

Garcés, Jorge Eduardo; Bozzolo, Guillermo

2011-01-01

We apply an atomistic modeling approach to deal with interfacial phenomena in high-density uranium fuels. The effects of Si, as additive to Al or as U-Mo-particles coating, on the behavior of the Al/U-Mo interface is modeled by using the Bozzolo-Ferrante-Smith (BFS) method for alloys. The basic experimental features characterizing the real system are identified, via simulations and atom-by-atom analysis. These include (1) the trend indicating formation of interfacial compounds, (2) much reduced diffusion of Al into U-Mo solid solution due to the high Si concentration, (3) Si depletion in the Al matrix, (4) an unexpected interaction between Mo and Simore » which inhibits Si diffusion to deeper layers in the U-Mo solid solution, and (5) the minimum amount of Si needed to perform as an effective diffusion barrier. Simulation results related to alternatives to Si dispersed in the Al matrix, such as the use of C coating of U-Mo particles or Zr instead of the Al matrix, are also shown. Recent experimental results confirmed early theoretical proposals, along the lines of the results reported in this work, showing that atomistic computational modeling could become a valuable tool to aid the experimental work in the development of nuclear fuels.« less

RECENT DEVELOPMENT IN TEM CHARACTERIZATION OF IRRADIATED RERTR FUELS

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

J. Gan; B.D. Miller; D.D. Keiser Jr.

2011-10-01

The recent development on TEM work of irradiated RERTR fuels includes microstructural characterization of the irradiated U-10Mo/alloy-6061 monolithic fuel plate, the RERTR-7 U-7Mo/Al-2Si and U-7Mo/Al-5Si dispersion fuel plates. It is the first time that a TEM sample of an irradiated nuclear fuel was prepared using the focused-ion-beam (FIB) lift-out technical at the Idaho National Laboratory. Multiple FIB TEM samples were prepared from the areas of interest in a SEM sample. The characterization was carried out using a 200kV TEM with a LaB6 filament. The three dimensional orderings of nanometer-sized fission gas bubbles are observed in the crystalline region of themore » U-Mo fuel. The co-existence of bubble superlattice and dislocations is evident. Detailed microstructural information along with composition analysis is obtained. The results and their implication on the performance of these fuels are discussed.« less

The irradiation behavior of atomized U-Mo alloy fuels at high temperature

NASA Astrophysics Data System (ADS)

Park, Jong-Man; Kim, Ki-Hwan; Kim, Chang-Kyu; Meyer, M. K.; Hofman, G. L.; Strain, R. V.

2001-04-01

Post-irradiation examinations of atomized U-10Mo, U-6Mo, and U-6Mo-1.7Os dispersion fuels from the RERTR-3 experiment irradiated in the Advanced Test Reactor (ATR) were carried out in order to investigate the fuel behavior of high uranium loading (8 gU/cc) at a high temperature (higher than 200°C). It was observed after about 40 at% BU that the U-Mo alloy fuels at a high temperature showed similar irradiation bubble morphologies compared to those at a lower temperature found in the RERTR-1 irradiation result, but there was a thick reaction layer with the aluminum matrix which was found to be greatly affected by the irradiation temperature and to a lesser degree by the fuel composition. In addition, the chemical analysis for the irradiated U-Mo fuels using the Electron Probe Micro Analysis (EPMA) method were conducted to investigate the compositional changes during the formation of the reaction product.

First-principles study of the surface properties of U-Mo system

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

Mei, Zhi-Gang; Liang, Linyun; Yacout, Abdellatif M.

U-Mo alloys are promising fuels for future high-performance research reactors with low enriched uranium. Surface properties, such as surface energy, are important inputs for mesoscale simulations (e.g., phase field method) of fission gas bubble behaviors in irradiated nuclear fuels. The lack of surface energies of U-Mo alloys prevents an accurate modeling of the morphology of gas bubbles and gas bubble-induced fuel swelling. To this end, we study the surface properties of U-Mo system, including bcc Mo, alpha-U, gamma-U, and gamma U-Mo alloys. All surfaces up to a maximum Miller index of three and two are calculated for cubic Mo andmore » gamma-U and non-cubic alpha-U, respectively. The equilibrium crystal shapes of bcc Mo, alpha-U and gamma-U are constructed using the calculated surface energies. The dominant surface orientations and the area fraction of each facet are determined from the constructed equilibrium crystal shape. The disordered gamma U-Mo alloys are simulated using the Special Quasirandom Structure method. The (1 1 0) and (1 0 0) surface energies of gamma U-7Mo and U-10Mo alloys are predicted to lie between those of gamma-U and bcc Mo, following a linear combination of the two constituents' surface energies. To better compare with future measurements of surface energies, the area fraction weighted surface energies of alpha-U, gamma-U and gamma U-7Mo and U-10Mo alloys are also predicted. (C) 2017 Published by Elsevier B.V.« less

Stability Study of the RERTR Fuel Microstructure

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

Jian Gan; Dennis Keiser; Brandon Miller

2014-04-01

The irradiation stability of the interaction phases at the interface of fuel and Al alloy matrix as well as the stability of the fission gas bubble superlattice is believed to be very important to the U-Mo fuel performance. In this paper the recent result from TEM characterization of Kr ion irradiated U-10Mo-5Zr alloy will be discussed. The focus will be on the phase stability of Mo2-Zr, a dominated second phase developed at the interface of U-10Mo and the Zr barrier in a monolithic fuel plate from fuel fabrication. The Kr ion irradiations were conducted at a temperature of 200 degreesmore » C to an ion fluence of 2.0E+16 ions/cm2. To investigate the thermal stability of the fission gas bubble superlattice, a key microstructural feature in both irradiated dispersion U-7Mo fuel and monolithic U-10Mo fuel, a FIB-TEM sample of the irradiated U-10Mo fuel (3.53E+21 fission/cm3) was used for a TEM in-situ heating experiment. The preliminary result showed extraordinary thermal stability of the fission gas bubble superlattice. The implication of the TEM observation from these two experiments on the fuel microstructural evolution under irradiation will be discussed.« less

Irradiation behavior of the interaction product of U-Mo fuel particle dispersion in an Al matrix

NASA Astrophysics Data System (ADS)

Kim, Yeon Soo; Hofman, G. L.

2012-06-01

Irradiation performance of U-Mo fuel particles dispersed in Al matrix is stable in terms of fuel swelling and is suitable for the conversion of research and test reactors from highly enriched uranium (HEU) to low enriched uranium (LEU). However, tests of the fuel at high temperatures and high burnups revealed obstacles caused by the interaction layers forming between the fuel particle and matrix. In some cases, fission gas filled pores grow and interconnect in the interdiffusion layer resulting in fuel plate failure. Postirradiation observations are made to examine the behavior of the interdiffusion layers. The interdiffusion layers show a fluid-like behavior characteristic of amorphous materials. In the amorphous interdiffusion layers, fission gas diffusivity is high and the material viscosity is low so that the fission gas pores readily form and grow. Based on the observations, a pore formation mechanism is proposed and potential remedies to suppress the pore growth are also introduced.

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).

A modelling study of the inter-diffusion layer formation in U-Mo/Al dispersion fuel plates at high power

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

Ye, B.; Hofman, G. L.; Leenaers, A.

Post irradiation examinations of full-size U-Mo/Al dispersion fuel plates fabricated with ZrN- or Sicoated U-Mo particles revealed that the reaction rate of irradiation-induced U-Mo-Al inter-diffusion, an important microstructural change impacting the performance of this type of fuel, is temperature and fission-rate dependent. In order to simulate the U-Mo/Al inter-diffusion layer (IL) growth behavior in full-size dispersion fuel plates, the existing IL growth correlation was modified with a temperaturedependent multiplication factor that transits around a threshold fission rate. In-pile irradiation data from four tests in the BR2 reactors, including FUTURE, E-FUTURE, SELEMIUM, and SELEMIUM-1a, were utilized to determine and validate themore » updated IL growth correlation. Irradiation behavior of the plates was simulated with the DART-2D computational code. The general agreement between the calculated and measured fuel meat swelling and constituent volume fractions as a function of fission density demonstrated the plausibility of the updated IL growth correlation. The simulation results also suggested the temperature dependence of the IL growth rate, similar to the temperature dependence of the intermixing rate in ion-irradiated bi-layer systems.« less

Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

DOE PAGES

Collette, R.; King, J.; Buesch, C.; ...

2016-04-01

The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends whenmore » comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. Here, the results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program.« less

Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

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

Collette, R.; King, J.; Buesch, C.

The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends whenmore » comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. Here, the results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program.« less

Thermal conductivity of fresh and irradiated U-Mo fuels

NASA Astrophysics Data System (ADS)

Huber, Tanja K.; Breitkreutz, Harald; Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.; Elgeti, Stefan; Reiter, Christian; Robinson, Adam. B.; Smith, Frances. N.; Wachs, Daniel. M.; Petry, Winfried

2018-05-01

The thermal conductivity of fresh and irradiated U-Mo dispersion and monolithic fuel has been investigated experimentally and compared to theoretical models. During in-pile irradiation, thermal conductivity of fresh dispersion fuel at a temperature of 150 °C decreased from 59 W/m·K to 18 W/m·K at a burn-up of 4.9·1021 f/cc and further to 9 W/m·K at a burn-up of 6.1·1021 f/cc. Fresh monolithic fuel has a considerably lower thermal conductivity of 15 W/m·K at a temperature of 150 °C and consequently its decrease during in-pile irradiation is less steep than for dispersion fuel. For a burn-up of 3.5·1021 f/cc of monolithic fuel, a thermal conductivity of 11 W/m·K at a temperature of 150 °C has been measured by Burkes et al. (2015). The difference of decrease for both fuels originates from effects in the matrix that occur during irradiation, like for dispersion fuel the gradual disappearance of the Al matrix with increased burn-up and the subsequent growth of an interaction layer (IDL) between the U-Mo fuel particle and Al matrix and subsequent matrix hardening. The growth of fission gas bubbles and the decomposition of the U-Mo crystal lattice also affect both dispersion and monolithic fuel.

Post irradiation analysis of RERTR-7A, 7B and RERTR-8 tests

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

Hofman, G.L.; Kim, Yeon Soo; Shevlyakov, G.V.

2008-07-15

Addition of 2 wt% or more of silicon in the Al matrix for U-Mo/Al dispersion fuel has proved to be effective in reducing interaction layer growth from the RERTR-7A test to a burnup of {approx}100 at% U-235 (LEU equivalent). The recent RERTR-8 test also showed the consistent results. In this paper, we present the post irradiation analysis results of these tests. A considerable number of monolithic fuel plates were irradiated in the RERTR-7A and RERTR-8 tests. The post irradiation results of these plates are also included. The RERTR-7B test was a lower burnup test with similar power to the RERTR-7A.more » In this test, dispersion fuel plates with U-7Mo-1Ti and U- 7Mo-2Zr in Al-5Si were irradiated. The post irradiation results of these plates are also covered. (author)« less

Pore growth in U-Mo/Al dispersion fuel

NASA Astrophysics Data System (ADS)

Kim, Yeon Soo; Jeong, G. Y.; Sohn, D.-S.; Jamison, L. M.

2016-09-01

U-Mo/Al dispersion fuel is currently under development in the DOE's Material Management and Minimization program to convert HEU-fueled research reactors to LEU-fueled reactors. In some demanding conditions in high-power and high-performance reactors, large pores form in the interaction layers between the U-Mo fuel particles and the Al matrix, which pose a potential to cause fuel failure. In this study, comprehension of the formation and growth of these pores was explored. As a product, a model to predict pore growth and porosity increase was developed. The model includes three major topics: fission gas release from the U-Mo and the IL to the pores, stress evolution in the fuel meat, and the effect of amorphous IL growth. Well-characterized in-pile data from reduced-size plates were used to fit the model parameters. A data set from full-sized plates, independent and distinctively different from those used to fit the model parameters, was used to examine the accuracy of the model. The model showed fair agreement with the measured data. The model suggested that the growth of the IL has a critical effect on pore growth, as both its material properties and energetics are favorable to pore formation. Therefore, one area of the current effort, focused on suppressing IL growth, appears to be on the right track to improve the performance of this fuel.

Interdiffusion and reactions between U-Mo and Zr at 650 °C as a function of time

NASA Astrophysics Data System (ADS)

Park, Y.; Keiser, D. D.; Sohn, Y. H.

2015-01-01

Development of monolithic U-Mo alloy fuel (typically U-10 wt.%Mo) for the Reduced Enrichment for Research and Test Reactors (RERTR) program entails a use of Zr diffusion barrier to eliminate the interdiffusion-reactions between the fuel alloy and Al-alloy cladding. The application of Zr barrier to the U-Mo fuel system requires a co-rolling process that utilizes a soaking temperature of 650 °C, which represents the highest temperature the fuel system is exposed to during both fuel manufacturing and reactor application. Therefore, in this study, development of phase constituents, microstructure and diffusion kinetics of U-10 wt.%Mo and Zr was examined using solid-to-solid diffusion couples annealed at 650 °C for 240, 480 and 720 h. Phase constituents and microstructural development were analyzed by scanning electron microscopy (SEM) and transmission electron microscopy (TEM). Concentration profiles were mapped as diffusion paths on the isothermal ternary phase diagram. Within the diffusion zone, single-phase layers of β-Zr and β-U were observed along with a discontinuous layer of Mo2Zr between the β-Zr and β-U layers. In the vicinity of Mo2Zr phase, islands of α-Zr phases were also found. In addition, acicular α-Zr and U6Zr3Mo phases were observed within the γ-U(Mo) terminal alloy. Growth rate of the interdiffusion-reaction zone was determined to be 7.75 (± 5.84) × 10-16 m2/s at 650 °C, however with an assumption of a certain incubation period.

Role of Si on the Diffusional Interactions Between U-Mo and Al-Si Alloys at 823 K (550 °C)

NASA Astrophysics Data System (ADS)

Perez, Emmanuel; Sohn, Yong-Ho; Keiser, Dennis D.

2013-01-01

U-Mo dispersions in Al-alloy matrix and monolithic fuels encased in Al-alloy are under development to fulfill the requirements for research and test reactors to use low-enriched molybdenum stabilized uranium alloy fuels. Significant interaction takes place between the U-Mo fuel and Al during manufacturing and in-reactor irradiation. The interaction products are Al-rich phases with physical and thermal characteristics that adversely affect fuel performance and result in premature failure. Detailed analysis of the interdiffusion and microstructural development of this system was carried through diffusion couples consisting of U-7 wt pct Mo, U-10 wt pct Mo and U-12 wt pct Mo in contact with pure Al, Al-2 wt pct Si, and Al-5 wt pct Si, annealed at 823 K (550 °C) for 1, 5 and 20 hours. Scanning electron microscopy and transmission electron microscopy were employed for the analysis. Diffusion couples consisting of U-Mo in contact with pure Al contained UAl3, UAl4, U6Mo4Al43, and UMo2Al20 phases. Additions of Si to the Al significantly reduced the thickness of the interdiffusion zone. The interdiffusion zones developed Al- and Si-enriched regions, whose locations and size depended on the Si and Mo concentrations in the terminal alloys. In these couples, the (U,Mo)(Al,Si)3 phase was observed throughout the interdiffusion zone, and the U6Mo4Al43 and UMo2Al20 phases were observed only where the Si concentrations were low.

Thermal conductivity of fresh and irradiated U-Mo fuels

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

Huber, Tanja K.; Breitkreutz, Harald; Burkes, Douglas E.

The thermal conductivity of fresh and irradiated U-Mo dispersion and monolithic fuel has been investigated experimentally and compared to theoretical models. During in-pile irradiation, the thermal conductivity of fresh dispersion fuel at a temperature of 150°C decreases from 59 W/m ·K down to 18  W/m ·K at a burn-up of 4.9 ·10 21 f/cc and further down to 9 W/m·K at a burn-up of 6.1·10 21 f/cc. Fresh monolithic fuel has a considerably lower thermal conductivity of 15 W/m·K at a temperature of 150 °C and consequently its decrease during in-pile irradiation is less steep as for the dispersion fuel. For a burn-up ofmore » 3.5·10 21 f /cc of monolithic fuel 11 W/m·K at a temperature of 150 °C has been measured by Burkes et al. The difference of the decrease of both fuels originates from effects in the matrix that occur during irradiation, like for dispersion fuel the gradual disappearance of the Al matrix with increasing burn-up and the subsequent growth of an interaction layer (IDL) between the U-Mo fuel particle and Al matrix and subsequent matrix hardening. The growth of fission gas bubbles and the decomposition of the U-Mo crystal lattice affects both dispersion and monolithic fuel.« less

A modelling study of the inter-diffusion layer formation in U-Mo/Al dispersion fuel plates at high power

NASA Astrophysics Data System (ADS)

Ye, B.; Hofman, G. L.; Leenaers, A.; Bergeron, A.; Kuzminov, V.; Van den Berghe, S.; Kim, Y. S.; Wallin, H.

2018-02-01

Post irradiation examinations of full-size U-Mo/Al dispersion fuel plates fabricated with ZrN- or Si- coated U-Mo particles revealed that the reaction rate of irradiation-induced U-Mo-Al inter-diffusion, an important microstructural change impacting the performance of this type of fuel, transited at a threshold temperature/fission rate. The existing inter-diffusion layer (IL) growth correlation, which does not describe the transition behavior of IL growth, was modified by applying a temperature-dependent multiplication factor that transits around a threshold fission rate. In-pile irradiation data from four tests in the BR2 reactors, including FUTURE, E-FUTURE, SELEMIUM, and SELEMIUM-1a, were utilized to determine and validate the updated IL growth correlation. Irradiation behavior of the plates was simulated with the DART-2D computational code. The general agreement between the calculated and measured fuel meat swelling and constituent volume fractions as a function of fission density demonstrated the plausibility of the updated IL growth correlation. The simulation results also suggested the temperature dependence of the IL growth rate, similar to the temperature dependence of the inter-mixing rate in ion-irradiated bi-layer systems.

Potential annealing treatments for tailoring the starting microstructure of low-enriched U-Mo dispersion fuels to optimize performance during irradiation

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Jue, Jan-Fong; Woolstenhulme, Nicolas E.; Ewh, Ashley

2011-12-01

Low-enriched uranium-molybdenum (U-Mo) alloy particles dispersed in aluminum alloy (e.g., dispersion fuels) are being developed for application in research and test reactors. To achieve the best performance of these fuels during irradiation, optimization of the starting microstructure may be required by utilizing a heat treatment that results in the formation of uniform, Si-rich interaction layers between the U-Mo particles and Al-Si matrix. These layers behave in a stable manner under certain irradiation conditions. To identify the optimum heat treatment for producing these kinds of layers in a dispersion fuel plate, a systematic annealing study has been performed using actual dispersion fuel samples, which were fabricated at relatively low temperatures to limit the growth of any interaction layers in the samples prior to controlled heat treatment. These samples had different Al matrices with varying Si contents and were annealed between 450 and 525 °C for up to 4 h. The samples were then characterized using scanning electron microscopy (SEM) to examine the thickness, composition, and uniformity of the interaction layers. Image analysis was performed to quantify various attributes of the dispersion fuel microstructures that related to the development of the interaction layers. The most uniform layers were observed to form in fuel samples that had an Al matrix with at least 4 wt.% Si and a heat treatment temperature of at least 475 °C.

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

Kim, Yeon Soo; Jeong, G. Y.; Sohn, D. -S.

U-Mo/Al dispersion fuel is currently under development in the DOE’s Material Management and Minimization program to convert HEU-fueled research reactors to LEU-fueled reactors. In some demanding conditions in high-power and high-performance reactors, large pores form in the interaction layers between the U-Mo fuel particles and the Al matrix, which pose a potential to cause fuel failure. In this study, comprehension of the formation and growth of these pores was explored. As a product, a model to predict pore growth and porosity increase was developed. Well-characterized in-pile data from reduced-size plates were used to fit the model parameters. A data setmore » of full-sized plates, independent and distinctively different from those used to fit the model parameters, was used to examine the accuracy of the model.« less

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

NASA Astrophysics Data System (ADS)

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

2014-05-01

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

Prediction of U-Mo dispersion nuclear fuels with Al-Si alloy using artificial neural network

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

Susmikanti, Mike, E-mail: mike@batan.go.id; Sulistyo, Jos, E-mail: soj@batan.go.id

2014-09-30

Dispersion nuclear fuels, consisting of U-Mo particles dispersed in an Al-Si matrix, are being developed as fuel for research reactors. The equilibrium relationship for a mixture component can be expressed in the phase diagram. It is important to analyze whether a mixture component is in equilibrium phase or another phase. The purpose of this research it is needed to built the model of the phase diagram, so the mixture component is in the stable or melting condition. Artificial neural network (ANN) is a modeling tool for processes involving multivariable non-linear relationships. The objective of the present work is to developmore » code based on artificial neural network models of system equilibrium relationship of U-Mo in Al-Si matrix. This model can be used for prediction of type of resulting mixture, and whether the point is on the equilibrium phase or in another phase region. The equilibrium model data for prediction and modeling generated from experimentally data. The artificial neural network with resilient backpropagation method was chosen to predict the dispersion of nuclear fuels U-Mo in Al-Si matrix. This developed code was built with some function in MATLAB. For simulations using ANN, the Levenberg-Marquardt method was also used for optimization. The artificial neural network is able to predict the equilibrium phase or in the phase region. The develop code based on artificial neural network models was built, for analyze equilibrium relationship of U-Mo in Al-Si matrix.« less

Detection of uranium and chemical state analysis of individual radioactive microparticles emitted from the Fukushima nuclear accident using multiple synchrotron radiation X-ray analyses.

PubMed

Abe, Yoshinari; Iizawa, Yushin; Terada, Yasuko; Adachi, Kouji; Igarashi, Yasuhito; Nakai, Izumi

2014-09-02

Synchrotron radiation (SR) X-ray microbeam analyses revealed the detailed chemical nature of radioactive aerosol microparticles emitted during the Fukushima Daiichi Nuclear Power Plant (FDNPP) accident, resulting in better understanding of what occurred in the plant during the early stages of the accident. Three spherical microparticles (∼2 μm, diameter) containing radioactive Cs were found in aerosol samples collected on March 14th and 15th, 2011, in Tsukuba, 172 km southwest of the FDNPP. SR-μ-X-ray fluorescence analysis detected the following 10 heavy elements in all three particles: Fe, Zn, Rb, Zr, Mo, Sn, Sb, Te, Cs, and Ba. In addition, U was found for the first time in two of the particles, further confirmed by U L-edge X-ray absorption near-edge structure (XANES) spectra, implying that U fuel and its fission products were contained in these particles along with radioactive Cs. These results strongly suggest that the FDNPP was damaged sufficiently to emit U fuel and fission products outside the containment vessel as aerosol particles. SR-μ-XANES spectra of Fe, Zn, Mo, and Sn K-edges for the individual particles revealed that they were present at high oxidation states, i.e., Fe(3+), Zn(2+), Mo(6+), and Sn(4+) in the glass matrix, confirmed by SR-μ-X-ray diffraction analysis. These radioactive materials in a glassy state may remain in the environment longer than those emitted as water-soluble radioactive Cs aerosol particles.

Detailed measurements of local thickness changes for U-7Mo dispersion fuel plates with Al-3.5Si matrix after irradiation at different powers in the RERTR-9B experiment

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Williams, Walter; Robinson, Adam; Wachs, Dan; Moore, Glenn; Crawford, Doug

2017-10-01

The Materials Management and Minimization program is developing fuel designs to replace highly enriched fuel with fuels of low enrichment. Swelling is an important irradiation behavior that needs to be well understood. Data from high resolution thickness measurements performed on U-7Mo dispersion fuel plates with Al-Si alloy matrices that were irradiated at high power is sparse. This paper reports the results of detailed thickness measurements performed on two dispersion fuel plates that were irradiated at relatively high power to high fission densities in the Advanced Test Reactor in the same RERTR-9B experiment. Both plates were irradiated to similar fission densities, but one was irradiated at a higher power than the other. The goal of this work is to identify any differences in the swelling behavior when fuel plates are irradiated at different powers to the same fission densities. Based on the results of detailed thickness measurments, more swelling occurs when a U-7Mo dispersion fuel with Al-3.5Si matrix is irradiated to a high fission density at high power compared to one irradiated at a lower power to high fission density.

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

Yim, J. S.; Tahk, Y. W.; Oh, J. Y.

In order to cope with global shortage of Mo-99 supplies and with growing demand of neutron transmutation doping, KJRR construction plan has been launched since April 2012 to provide self-sufficiency of domestic RI demand, and to extend Si doping capacity for power device market growth. Through comprehensive surveillance of the fuels in-reactor behavior, KAERI has selected the fuel meat of U-7%Mo dispersion in an aluminum matrix with 5wt%Si for the KJRR fuel. As part of the efforts for fuel licensing and qualification of the KJRR fuel, an LTA irradiation test at the ATR started from November 2015 was successfully completedmore » by reaching at 219 EFPD in the end of February 2017. Together with the results of HAMP-1 already completed irradiation and PIE, the successful irradiation of the LTA also demonstrates the fuel integrity under more rigorous conditions than the KJRR operation conditions. This paper updates the current status of the KJRR U7Mo (8 g-U/cm3) LTA irradiation and PIE plan up to date as of February 2017.« less

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

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

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

2008-07-15

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

Effects of heat treatment on U-Mo fuel foils with a zirconium diffusion barrier

NASA Astrophysics Data System (ADS)

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

2015-05-01

A monolith fuel design based on U-Mo alloy has been selected as the fuel type for conversion of the United States' high performance research reactors (HPRRs) from highly enriched uranium (HEU) to low-enriched uranium (LEU). In this fuel design, a thin layer of zirconium is used to eliminate the direct interaction between the U-Mo fuel meat and the aluminum-alloy cladding during irradiation. The co-rolling process used to bond the Zr barrier layer to the U-Mo foil during fabrication alters the microstructure of both the U-10Mo fuel meat and the U-Mo/Zr interface. This work studied the effects of post-rolling annealing treatment on the microstructure of the co-rolled U-Mo fuel meat and the U-Mo/Zr interaction layer. Microscopic characterization shows that the grain size of U-Mo fuel meat increases with the annealing temperature, as expected. The grain sizes were ∼9, ∼13, and ∼20 μm for annealing temperature of 650, 750, and 850 °C, respectively. No abnormal grain growth was observed. The U-Mo/Zr interaction-layer thickness increased with the annealing temperature with an Arrhenius constant for growth of 184 kJ/mole, consistent with a previous diffusion-couple study. The interaction layer thickness was 3.2 ± 0.5 μm, 11.1 ± 2.1 μm, 27.1 ± 0.9 μm for annealing temperature of 650, 750, to 850 °C, respectively. The homogeneity of Mo improves with post rolling annealing temperature and with U-Mo coupon homogenization. The phases in the Zr/U-Mo interaction layer produced by co-rolling, however, differ from those reported in the previous diffusion couple studies.

Micro-structural study and Rietveld analysis of fast reactor fuels: U-Mo fuels

NASA Astrophysics Data System (ADS)

Chakraborty, S.; Choudhuri, G.; Banerjee, J.; Agarwal, Renu; Khan, K. B.; Kumar, Arun

2015-12-01

U-Mo alloys are the candidate fuels for both research reactors and fast breeder reactors. In-reactor performance of the fuel depends on the microstructural stability and thermal properties of the fuel. To improve the fuel performance, alloying elements viz. Zr, Mo, Nb, Ti and fissium are added in the fuel. The first reactor fuels are normally prepared by injection casting. The objective of this work is to compare microstructure, phase-fields and hardness of as-cast four different U-Mo alloy (2, 5, 10 and 33 at.% Mo) fuels with the equilibrium microstructure of the alloys. Scanning electron microscope with energy dispersive spectrometer and optical microscope have been used to characterize the morphology of the as-cast and annealed alloys. The monoclinic α'' phase in as-cast U-10 at.% Mo alloy has been characterized through Rietveld analysis. A comparison of metallographic and Rietveld analysis of as-cast (dendritic microstructure) and annealed U-33 at.% Mo alloy, corresponding to intermetallic compound, has been reported here for the first time. This study will provide in depth understanding of microstructural and phase evolution of U-Mo alloys as fast reactor fuel.

Irradiation testing of high density uranium alloy dispersion fuels

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

Hayes, S.L.; Trybus, C.L.; Meyer, M.K.

1997-10-01

Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 microplates. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U-10Mo-0.05Sn, U{sub 2}Mo, or U{sub 3}Si{sub 2}. These experiments will be discharged at peak fuel burnups ofmore » 40% and 80%. Of particular interest is the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions.« less

Phase development in a U-7 wt.% Mo vs. Al-7 wt.% Ge diffusion couple

NASA Astrophysics Data System (ADS)

Perez, E.; Keiser, D. D.; Sohn, Y. H.

2013-10-01

Fuel development for the Reduced Enrichment for Research and Test Reactors (RERTR) program has demonstrated that U-Mo alloys in contact with Al develop interaction regions with phases that have poor irradiation behavior. The addition of Si to the Al has been considered with positive results. In this study, compositional modification is considered by replacing Si with Ge to determine the effect on the phase development in the system. The microstructural and phase development of a diffusion couple of U-7 wt.% Mo in contact with Al-7 wt.% Ge was examined by transmission electron microscopy, scanning electron microscopy and energy dispersive spectroscopy. The interdiffusion zone developed a microstructure that included the cubic-UGe3 phase and amorphous phases. The UGe3 phase was observed with and without Mo and Al solid solution developing a (U,Mo)(Al,Ge)3 phase.

An interdiffusional model for prediction of the interaction layer growth in the system uranium molybdenum/aluminum

NASA Astrophysics Data System (ADS)

Soba, A.; Denis, A.

2007-03-01

The codes PLACA and DPLACA, elaborated in this working group, simulate the behavior of a plate-type fuel containing in its core a foil of monolithic or dispersed fissile material, respectively, under normal operation conditions of a research reactor. Dispersion fuels usually consist of ceramic particles of a uranium compound in a high thermal conductivity matrix. The use of particles of a U-Mo alloy in a matrix of Al requires especially devoted subroutines able to simulate the growth of the interaction layer that develops between the particles and the matrix. A model is presented in this work that gives account of these particular phenomena. It is based on the assumption that diffusion of U and Al through the layer is the rate-determining step. Two moving interfaces separate the growing reaction layer from the original phases. The kinetics of these boundaries are solved as Stefan problems. In order to test the model and the associated code, some previous, simpler problems corresponding to similar systems for which analytical solutions or experimental data are known were simulated. Experiments performed with planar U-Mo/Al diffusion couples are reported in the literature, which purpose is to obtain information on the system parameters. These experiments were simulated with PLACA. Results of experiments performed with U-Mo particles disperse in Al either without or with irradiation, published in the open literature were simulated with DPLACA. A satisfactory prediction of the whole reaction layer thickness and of the individual fractions corresponding to alloy and matrix consumption was obtained.

Effects of thermal treatment on the co-rolled U-Mo fuel foils

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

Dennis D. Keiser, Jr.; Tammy L. Trowbridge; Cynthia R. Breckenridge

2014-11-01

A monolithic fuel type is being developed to convert US high performance research and test reactors such as Advanced Test Reactor (ATR) at Idaho National Laboratory from highly enriched uranium (HEU) to low-enriched uranium (LEU). The interaction between the cladding and the U-Mo fuel meat during fuel fabrication and irradiation is known to have negative impacts on fuel performance, such as mechanical integrity and dimensional stability. In order to eliminate/minimize the direct interaction between cladding and fuel meat, a thin zirconium diffusion barrier was introduced between the cladding and U-Mo fuel meat through a co-rolling process. A complex interface betweenmore » the zirconium and U-Mo was developed during the co-rolling process. A predictable interface between zirconium and U-Mo is critical to achieve good fuel performance since the interfaces can be the weakest link in the monolithic fuel system. A post co-rolling annealing treatment is expected to create a well-controlled interface between zirconium and U-Mo. A systematic study utilizing post co-rolling annealing treatment has been carried out. Based on microscopy results, the impacts of the annealing treatment on the interface between zirconium and U-Mo will be presented and an optima annealing treatment schedule will be suggested. The effects of the annealing treatment on the fuel performance will also be discussed.« less

Neutronic performance of high-density LEU fuels in water-moderated and water-reflected research reactors

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

Bretscher, M.M.; Matos, J.E.

At the Reduced Enrichment for Research and Test Reactors (RERTR) meeting in September 1994, Durand reported that the maximum uranium loading attainable with U{sub 3}Si{sub 2} fuel is about 6.0 g U/cm{sup 3}. The French Commissariat a l`Energie Atomique (CEA) plan to perform irradiation tests with 5 plates at this loading. Compagnie pour L`Etude et La Realisation de Combustibles Atomiques (CERCA) has also fabricated a few uranium nitride (UN) plates with a uranium density in the fuel meat of 7.0 g/cm{sup 3} and found that UN is compatible with the aluminum matrix at temperatures below 500 C. High density dispersionmore » fuels proposed for development include U-Zr(4 wt%)-Nb(2 wt%), U-Mo(5 wt%), and U-Mo(9 wt%). The purpose of this note is to examine the relative neutronic behavior of these high density fuels in a typical light water-reflected and water-moderated MTR-type research reactor. The results show that a dispersion of the U-Zr-Nb alloy has the most favorable neutronic properties and offers the potential for uranium densities greater than 8.0 g/cm{sup 3}. On the other hand, UN is the least reactive fuel because of the relatively large {sup 14}N(n,p) cross section. For a fixed value of k{sub eff}, the required {sup 235}U loading per fuel element is least for the U-Zr-Nb fuel and steadily increases for the U-Mo(5%), U-Mo(9%), and UN fuels. Because of volume fraction limitations, the UO{sub 2} dispersions are only useful for uranium densities below 5.0 g/cm{sup 3}. In this density range, however, UO{sub 2} is more reactive than U{sub 3}Si{sub 2}.« less

IRRADIATION TESTING OF THE RERTR FUEL MINIPLATES WITH BURNABLE ABSORBERS IN THE ADVANCED TEST REACTOR

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

I. Glagolenko; D. Wachs; N. Woolstenhulme

2010-10-01

Based on the results of the reactor physics assessment, conversion of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) can be potentially accomplished in two ways, by either using U-10Mo monolithic or U-7Mo dispersion type plates in the ATR fuel element. Both designs, however, would require incorporation of the burnable absorber in several plates of the fuel element to compensate for the excess reactivity and to flatten the radial power profile. Several different types of burnable absorbers were considered initially, but only borated compounds, such as B4C, ZrB2 and Al-B alloys, were selected for testing primarily duemore » to the length of the ATR fuel cycle and fuel manufacturing constraints. To assess and compare irradiation performance of the U-Mo fuels with different burnable absorbers we have designed and manufactured 28 RERTR miniplates (20 fueled and 8 non-fueled) containing fore-mentioned borated compounds. These miniplates will be tested in the ATR as part of the RERTR-13 experiment, which is described in this paper. Detailed plate design, compositions and irradiations conditions are discussed.« less

Effect of in-pile degradation of the meat thermal conductivity on the maximum temperature of the plate-type U-Mo dispersion fuels

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

Pavel G. Medvedev

2009-11-01

Effect of in-pile degradation of thermal conductivity on the maximum temperature of the plate-type research reactor fuels has been assessed using the steady-state heat conduction equation and assuming convection cooling. It was found that due to very low meat thickness, characteristic for this type of fuel, the effect of thermal conductivity degradation on the maximum fuel temperature is minor. For example, the fuel plate featuring 0.635 mm thick meat operating at heat flux of 600 W/cm2 would experience only a 20oC temperature rise if the meat thermal conductivity degrades from 0.8 W/cm-s to 0.3 W/cm-s. While degradation of meat thermalmore » conductivity in dispersion-type U-Mo fuel can be very substantial due to formation of interaction layer between the particles and the matrix, and development of fission gas filled porosity, this simple analysis demonstrates that this phenomenon is unlikely to significantly affect the temperature-based safety margin of the fuel during normal operation.« less

Microstructure of as-fabricated UMo/Al(Si) plates prepared with ground and atomized powder

NASA Astrophysics Data System (ADS)

Jungwirth, R.; Palancher, H.; Bonnin, A.; Bertrand-Drira, C.; Borca, C.; Honkimäki, V.; Jarousse, C.; Stepnik, B.; Park, S.-H.; Iltis, X.; Schmahl, W. W.; Petry, W.

2013-07-01

UMo-Al based fuel plates prepared with ground U8wt%Mo, ground U8wt%MoX (X = 1 wt%Pt, 1 wt%Ti, 1.5 wt%Nb or 3 wt%Nb) and atomized U7wt%Mo have been examined. The first finding is that that during the fuel plate production the metastable γ-UMo phases partly decomposed into two different γ-UMo phases, U2Mo and α'-U in ground powder or α″-U in atomized powder. Alloying small amounts of a third element to the UMo had no measurable effect on the stability of the γ-UMo phase. Second, the addition of some Si inside the Al matrix and the presence of oxide layers in ground and atomized samples is studied. In the case with at least 2 wt%Si inside the matrix a Silicon rich layer (SiRL) forms at the interface between the UMo and the Al during the fuel plate production. The SiRL forms more easily when an Al-Si alloy matrix - which is characterized by Si precipitates with a diameter ⩽1 μm - is used than when an Al-Si mixed powder matrix - which is characterized by Si particles with some μm diameter - is used. The presence of an oxide layer on the surface of the UMo particles hinders the formation of the SiRL. Addition of some Si into the Al matrix [7-11]. Application of a protective barrier at the UMo/Al interface by oxidizing the UMo powder [7,12]. Increase of the Mo content or use of UMo alloys with ternary element addition X (e.g. X = Nb, Ti, Pt) to stabilize the γ-UMo with respect to α-U or to control the UMo-Al interaction layer kinetics [9,12-24]. Use of ground UMo powder instead of atomized UMo powder [10,25] The points 1-3 are to limit the formation of the undesired UMo/Al layer. Especially the addition of Si into the matrix has been suggested [3,7,8,10,11,26,27]. It has been often mentioned that Silicon is efficient in reducing the Uranium-Aluminum diffusion kinetics since Si shows a higher chemical affinity to U than Al to U. Si suppresses the formation of brittle UAl4 which causes a huge swelling during the irradiation. Furthermore it enhances the formation of more stable UAl3 within the diffusion layer [14]. In addition, Si will not notably influence the reactor neutronics due to its low absorption cross section for thermal neutrons of σabs = 0.24 barn. Aluminum has σabs = 0.23 barn.Williams [28], Bierlein [29], Green [30] and de Luca [31] showed the first time in the 1950s that alloying Aluminum with some Silicon reduces the Uranium-Aluminum diffusion kinetics in can-type fuel elements. However, up to now uncertainties remained about the most promising Si concentration and the involved mechanisms.Ground powder - possibility 4 - introduces a high density of defects like dislocations, oxide layers and impurities into UMo grains. Fuel prepared with this kind of powder exhibits a larger porosity. It may also be combined with an AlSi matrix. As a consequence, the degree of swelling due to high-burn up is reduced compared to fuel with atomized powder [5,6,25].This study focuses on the metallurgical characterization of as-fabricated samples prepared with ground UMo and UMoX (X = Ti, Nb, Pt) powders and atomized UMo powder. The influence of some Si into the Al matrix and the presence of oxide layers on the UMo is discussed. Details of the differences of samples with ground UMo from atomized UMo will be discussed.The examined samples originate from non-irradiated spare fuel plates from the IRIS-TUM irradiation campaign [5,6]. The samples containing ground UMoX powders and atomized UMo powders with Si addition into the matrix have been produced for this study [32]. Powder mixing: The UMo powder is mixed with Al powder. Compact production: UMo-Al powder is poured into a mould and undergoes compaction under large force. Plate-processing: An AlFeNi frame is placed on an AlFeNi plate and the UMo-Al compact is placed into the frame. Afterwards it is covered with a second AlFeNi plate. This assembly is hot-rolled to reduce the total thickness to 1.4 mm. Subsequently, a blister test (1-2 h at 400-450 °C) ensures that the fuelplate is sealed. After this step, the UMo particles are tightly covered with Al as shown in Fig. 1. To access the meat layer, small samples have been cut from the fuel plates. The AlFeNi cladding has been removed using abrasive paper and diamond polishing paste. Cross sections were prepared from each sample and examined using SEM/EDX and XRD. Laboratory scale XRD Laboratory sealed-tube XRD on a STOE-STADIP diffractometer equipped with an incident beam focusing monochromator and used in reflection geometry with respect to the sample. MoK-α radiation has been used. Details on the systems used can be found in [39]. mu;-XRD using micro-focused synchrotron radiation at the Swiss Light Source μ-XAS beamline (PSI, Switzerland). At SLS, the beam size was 3 × 3 μm2, the energy was 19.7 keV. Further details on the experimental procedure can be found in [40]. Only very small sample volumes are probed with this technique, therefore the results may not be representative for the whole miniplate. The standard deviation of the lattice parameters obtained with this method is ±0.01 Å in case not different given. High-energy XRD (HE-XRD) in transmission mode using synchrotron radiation at the "High Energy Diffraction and Scattering Beamline ID15B" of ESRF. An X-ray energy of 87 keV has been used, the beam size was 0.3 × 0.3 mm2. Details on the experimental procedure are presented in [41,42]. It was possible to determine the average mass fractions of the phases present inside the sample using this technique. The standard deviation of the lattice parameters obtained with this method is ±0.001 Å in case not different given. laser granulometry to determine the size distribution of the particles, XRD for phase identification. Granulometry measurements showed that a significant amount of very fine particles of a few μm to 10 μm size are present in the first class of powder.In both cases, laboratory XRD analyses evidenced only two phases: γ-UMo and UO2. In contrast to observations on the final fuel plates, there are no signs of α-U. Comparing XRD data of atomized UMo powder (taken form the IRIS4 experiment) and ground UMo powder with almost the same Mo content, the peaks are broader in XRD patterns of ground UMo and there is a higher background [44]. This points that the lattice structure of the UMo inside the ground powder is strongly disordered during the grinding process.Complementary investigations were performed in these ground UMo powder samples using HE-XRD. The obtained data can therefore directly be compared to those measured on pre-oxidized atomized UMo powders [45] and IRIS-TUM fuel plates [41]. For both powder samples the γ-UMo lattice constant has been estimated to 3.433 ± 0.002 Å which corresponds to about 7.2 wt% for Mo in the alloy according to Dwight's law [46]. The existence of two UMo phases inside these ground particles (as in atomized case) could not be investigated because of the huge peak broadening (presence of micro distortions). Whatever the sample granulometry, the analysis of the HE-XRD data showed a non-negligible nitride contamination in ground powders (see Fig. 2). Two uranium nitride phases are indeed found in these samples: UN and U2N3+x[47]. Note that the presence of UN has also been found in the as-fabricated plates. These results confirm the high reactivity of UMo with both Oxygen and Nitrogen in the grinding conditions. As a comparison for temperatures in the 200-250 °C range, it seems that UNx phases are more difficult to grow: they were not present in outer layers obtained by heat treatment under air on atomized particles [45]. Finally it can be seen in Table 3 that the weight fractions of UO2 and U2N3+x phases are lower in the sample with larger UMo particles. This suggests the existence of an oxide, nitride outer shell around UMo ground particle with thickness that does not strongly evolve with particle size. This constant outer shell thickness has also been found in pre-oxidized atomized powders [45].The UMoX powder used for the samples MAFIA-I-18 - MAFIA-I-21 has not been investigated prior to plate fabrication. However, since the grinding process is essentially the same as for the pure UMo powder, similar characteristics are assumed. Thin oxide layers with a thickness ⩽1 μm on some of the particles that were not intentionally oxidized. Although the UMo powder was stored and handled under an inert atmosphere over the whole production process, some residual oxygen has reacted with the UMo. Already this thin oxide layers exhibits cracks (Fig. 5). Thicker oxide layers with a thickness up to 5 μm on the UMo particles that were oxidized purposely. This kind of oxide layer is very brittle and shows large cracks (Fig. 6). The oxidized UMo particles tend to detach with the matrix as gaps between the UMo particles and the oxide layer could be observed (Fig. 6). This is most obvious at spots where a UMo particle has been pulled out during polishing. A part of the oxide layer remained inside the resulting hole (Fig. 7). Atomized UMo powder 2 wt%Si in Al matrix, alloyed AlSi 2 wt%Si in Al matrix, mixed AlSi 5 wt%Si in Al matrix, mixed AlSi 7 wt%Si in Al matrix, mixed AlSi Ground UMo powder 2 wt%Si in Al matrix, alloyed AlSi The influence of an oxide layer around the UMo particles on the formation of the SiRL during fuel plate production is further discussed. The growth of a Si rich layer surrounding the UMo particles in the 2 wt%Si alloyed powder (oxidized UMo), as well as the 5 wt% and 7 wt%Si mixed powder (non-oxidized UMo) during production of the miniplates. The presence of Si precipitates in the Al-matrix (large precipitates in case of mixture, small si particles in alloy). No oxide layer: If no oxide layer is present around the UMo particles a homogeneous SiRL grows at the interface UMo-Al (Fig. 15a). Brittle oxide layer: An oxide layer is present around the UMo particles, the SiRL grows always between the UMo particle and the oxide layer (Fig. 15b). In this case the the SiRL is thin and not homogeneous. As presumed by Ripert et al. [7] it is essential that the oxide layer reveals cracks perpendicular the UMo particle surface to make path for the Si diffusion. Dense oxide layer: In case of a thin (≈1 μm) but compact oxide layer no SiRL is formed even at high Si concentrations inside the matrix (Fig. 15c). The observed effects are pronounced when the thickness of the oxide layer is increased, as shown in Fig. 16: UMo particles covered with a thicker oxide layer (>1 μm) inside an Aluminum matrix with 5 wt%Si (mixed Al-Si powder). The oxide layer is dense at the left side of the particle, no Si can be found there (Fig. 16a). In contrast, the brittle and cracked oxide layer on the right side made path for a Si diffusion but the SiRL is thinner than in the sites where the UMo particle is not covered with an oxide layer. EDX maps at different positions of the sample showed that in general no SiRL forms around UMo particles covered by oxide layers with a thickness greater than 1 μm (Fig. 16b). This behavior is identical for the samples with 5 wt%Si and 7 wt%Si added to the Aluminum matrix (mixed Al-Si powder). Obviously the presence of a (dense) oxide layer hampers the formation of a SiRL. different UXSiY phases with strongly overlapping peaks can be found in the SiRL, these phases are characterized by small sizes of the crystallites (a few tens of nanometers) and/or cell parameter gradients. Two different crystallographic phases have been usually identified: U(Al,Si)3 displaying a small lattice parameter of a0 = 4.16 Å. This indicates that about 40% of the Al lattice sites are occupied by Si atoms. The second phase is isostructural to U3Si54 with a different lattice parameter [59-61]. Although the U-Si-Al phase diagram contains a variety of phases, none of the phases reported in literature [62] could be used to fully refine the measured diagram. Therefore, three different hypotheses are suggested to explain the occurence of this unknown phase: The observed compound consists of two phases: Conventional U3Si5 and USi2, as has been suggested by the authors before [58]. However, only one literature source (Brown et al.) describing the occurrence of USi2 below 450 °C could be found [63,64]. Furthermore, it has not been possible to reproduce the experiments described by Brown et al. Therefore, this hypothesis remains doubtful [59]. The observed phase may be a new unknown phase. For example, a cubic phase with lattice constant a0 = 3.96 Å can be used to refine the observed peaks. This hypothesis can neither be confirmed nor refused based on the existing data. The observed phase can be a U3Si5 variant containing Mo and/or Al atoms. This hypothesis is supported by the authors. Hence in the following sections this structure will be denoted as U3Si5. No traces of SiRL phases are found inside the sample with 2 wt%Si mixed-powder matrix (MAFIA-I-3), all the Si remained inside the matrix. A SiRL is present inside the samples with 2.1 wt%Si alloyed powder matrix (MAFIA-I-4) and 5 wt%Si (MAFIA-I-5) and 7 wt%Si (MAFIA-I-7) mixed powder matrix. However, between 76% and 96% of the Si remained inside the matrix in form of precipitates or Si particles. The SiRL is formed readily when the Si is present inside the matrix in form of precipitates (i.e. Al-Si alloy matrix, MAFIA-I-4 and IRIS-TUM 8502) compared to particles (i.e. Al-Si mixed powder matrix, MAFIA-I-3, MAFIA-I-5 and MAFIA-I-7). This behavior can best be observed on the sample prepared with ground powder and with 2.1 wt%Si alloyed powder matrix (IRIS-TUM-8502): The matrix contains no Si, only SiRL phases are found. Since the sample with 5 wt%Si mixed powder matrix (MAFIA-I-5) has the lowest SiRL fraction but by far the highest UO2 content, it is concluded that the presence of UO2 around the UMo kernels tends to hamper the formation of a SiRL. UMo/Al samples prepared with ground powder contain irregularly shaped UMo kernels. They are in general oxidized and also contain oxide stringers. These samples have a high porosity content of around 8 vol%. In contrast, UMo/Al samples prepared with atomized powder contain spherical UMo kernels. Only the surface of the UMo kernels is oxidized in some cases. Thick oxide layers must be grown intentionally while thinner layers are the result of oxidation during the whole process. The oxide layer is in general brittle and exhibits cracks. The Uranium-oxide content of all examined samples (atomized and ground) varies between 2 and 13 wt%. gamma;-UMo present in the fresh UMo powder destabilizes to transform to an α-U-like phase, U2Mo, and two γ-UMo phases with different Mo content during the fuel plate production. For ground powder, α-U content varies in 28-38 wt%, for atomized powder in 11-14 wt%. The degree of γ-phase destabilization is therefore higher for ground powder. Ternary addition of Nb, Ti or Pt to the UMo did not impact the extent of decomposition. The γ-phase decomposition in the atomized and ground powder does not follow the expected in the U8wt%Mo TTT diagram between 400 and 450 °C [41]. According to Repas et al. [65], the route is γ-UMoa → γ-UMob + α-U → γ-UMoc+α-U + U2Mo . γ-UMoa,b,c differ in the Mo content where γ-UMoa has the lowest and γ-UMoc has the highest Mo content. We observe a new route of decomposition of ground powder into two different γ-UMo phases. One of them has approximately the original Mo content and the other has a higher Mo content. Further U2Mo and a phase with deformed lattice parameters compared to pure α-U have been observed. The latter is known as α' in literature.For atomized powder, also two different γ-UMo phases and traces of U2Mo have been found. However, a different α-U like phase has been identified: α″ [41,53-55].Repas et al. used as cast samples that have been examined with conventional XRD and different metallographic methods [65]. The difference to our data can be explained by the superior resolution of the here used HE-XRD diffraction. Most probably, conventional lab X-ray sourcces could not resolve fine differences in the lattice parameters of α-U and may not enable to separate two γ-UMo phase. To overcome this uncertainty it is highly desirable to examine the TTT diagram of UMo with high resolution. When Si is added into the matrix - by using alloyed Al-Si powder as a matrix or blending Al and Si powder - in general a SiRL is formed at the interface between the UMo and the Al matrix. An exception can be found in MAFIA-I-3 in which the overall Si content was to low to form a SiRL. The SiRL consists of U(Al,Si)3 and U3Si5. The SiRL forms less readily in case of mixed Al-Si than in case of alloyed Al-Si powder. In the latter case (MAFIA-I-4), a Si depleted zone has been observed around the UMo particles. For ground powder in combination with an Al-Si alloyed matrix, the entire Si from the matrix has reacted with the UMo forming SiRL phases. The presence of a dense oxide layer around the UMo kernels can prevent the formation of a SiRL. However, as soon as the oxide layer is cracked a SiRL forms between the UMo and the oxide layer. A dense oxide layer isolates the UMo from the Si inside the matrix and occurring cracks make path for the diffusion of Si towards the UMo. U3Si 5 is also called USi2-x or USi1.66 in literature.

U-Mo Plate Blister Anneal Interim Report

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

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

2010-10-01

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

Cubic γ-phase U-Mo alloys synthesized by splat-cooling

NASA Astrophysics Data System (ADS)

Kim-Ngan, Nhu-T. H.; Tkach, I.; Mašková, S.; Havela, L.; Warren, A.; Scott, T.

2013-09-01

U-Mo alloys are the most promising materials fulfilling the requirements of using low enriched uranium (LEU) fuel in research reactors. From a fundamental standpoint, it is of interest to determine the basic thermodynamic properties of the cubic γ-phase U-Mo alloys. We focus our attention on the use of Mo doping together with ultrafast cooling (with high cooling rates ⩾106 K s-1), which helps to maintain the cubic γ-phase in U-Mo system to low temperatures and on determination of the low-temperature properties of these γ-U alloys. Using a splat cooling method it has been possible to maintain some fraction of the high-temperature γ-phase at room temperature in pure uranium. U-13 at.% Mo splat clearly exhibits the pure γ-phase structure. All the splats become superconducting with Tc in the range from 1.24 K (pure U splat) to 2.11 K (U-15 at.% Mo). The γ-phase in U-Mo alloys undergoes eutectoid decomposition to form equilibrium phases of orthorhombic α-uranium and tetragonal γ‧-phase upon annealing at 500 °C, while annealing at 800 °C has stabilized the initial γ phase. The α-U easily absorbs a large amount of hydrogen (UH3 hydride), while the cubic bcc phase does not absorb any detectable amount of hydrogen at pressures below 1 bar and at room temperature. At 80 bar, the U-15 at.% Mo splat becomes powder consisting of elongated particles of 1-2 mm, revealing amorphous state.

Diffusion Barrier Selection from Refractory Metals (Zr, Mo and Nb) via Interdiffusion Investigation for U-Mo RERTR Fuel Alloy

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

K. Huang; C. Kammerer; D. D. Keiser, Jr.

2014-04-01

U-Mo alloys are being developed as low enrichment monolithic fuel under the Reduced Enrichment for Research and Test Reactor (RERTR) Program. Diffusional interactions between the U-Mo fuel alloy and Al-alloy cladding within the monolithic fuel plate construct necessitate incorporation of a barrier layer. Fundamentally, a diffusion barrier candidate must have good thermal conductivity, high melting point, minimal metallurgical interaction, and good irradiation performance. Refractory metals, Zr, Mo, and Nb are considered based on their physical properties, and the diffusion behavior must be carefully examined first with U-Mo fuel alloy. Solid-to-solid U-10wt.%Mo vs. Mo, Zr, or Nb diffusion couples were assembledmore » and annealed at 600, 700, 800, 900 and 1000 degrees C for various times. The interdiffusion microstructures and chemical composition were examined via scanning electron microscopy and electron probe microanalysis, respectively. For all three systems, the growth rate of interdiffusion zone were calculated at 1000, 900 and 800 degrees C under the assumption of parabolic growth, and calculated for lower temperature of 700, 600 and 500 degrees C according to Arrhenius relationship. The growth rate was determined to be about 10 3 times slower for Zr, 10 5 times slower for Mo and 10 6 times slower for Nb, than the growth rates reported for the interaction between the U-Mo fuel alloy and pure Al or Al-Si cladding alloys. Zr, however was selected as the barrier metal due to a concern for thermo- mechanical behavior of UMo/Nb interface observed from diffusion couples, and for ductile-to-brittle transition of Mo near room temperature.« less

Surface engineering of low enriched uranium-molybdenum

NASA Astrophysics Data System (ADS)

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

2013-09-01

Recent attempts to qualify the LEU(Mo) dispersion plate fuel with Si addition to the Al matrix up to high power and burn-up have not yet been successful due to unacceptable fuel plate swelling at a local burn-up above 60% 235U. The root cause of the failures is clearly related directly to the formation of the U(Mo)-Al(Si) interaction layer. Excessive formation of these layers around the fuel kernels severely weakens the local mechanical integrity and eventually leads to pillowing of the plate. In 2008, SCK·CEN has launched the SELENIUM U(Mo) dispersion fuel development project in an attempt to find an alternative way to reduce the interaction between U(Mo) fuel kernels and the Al matrix to a significantly low level: by applying a coating on the U(Mo) kernels. Two fuel plates containing 8gU/cc U(Mo) coated with respectively 600 nm Si and 1000 nm ZrN in a pure Al matrix were manufactured. These plates were irradiated in the BR2 reactor up to a maximum heat flux of 470 W/cm2 until a maximum local burn-up of approximately 70% 235U (˜50% plate average) was reached. Awaiting the PIE results, the advantages of applying a coating are discussed in this paper through annealing experiments and TRIM (the Transport of Ions in Matter) calculations.

Microstructural analysis of as-processed U-10 wt.%Mo monolithic fuel plate in AA6061 matrix with Zr diffusion barrier

NASA Astrophysics Data System (ADS)

Perez, E.; Yao, B.; Keiser, D. D., Jr.; Sohn, Y. H.

2010-07-01

For higher U-loading in low-enriched U-10 wt.%Mo fuels, monolithic fuel plate clad in AA6061 is being developed as a part of Reduced Enrichment for Research and Test Reactor (RERTR) program. This paper reports the first characterization results from a monolithic U-10 wt.%Mo fuel plate with a Zr diffusion barrier that was fabricated as part of a plate fabrication campaign for irradiation testing in the Advanced Test Reactor (ATR). Both scanning and transmission electron microscopy (SEM and TEM) were employed for analysis. At the interface between the Zr barrier and U-10 wt.%Mo, going from Zr to U(Mo), UZr 2, γ-UZr, Zr solid-solution and Mo 2Zr phases were observed. The interface between AA6061 cladding and Zr barrier plate consisted of four layers, going from Al to Zr, (Al, Si) 2Zr, (Al, Si)Zr 3 (Al, Si) 3Zr, and AlSi 4Zr 5. Irradiation behavior of these intermetallic phases is discussed based on their constituents. Characterization of as-fabricated phase constituents and microstructure would help understand the irradiation behavior of these fuel plates, interpret post-irradiation examination, and optimize the processing parameters of monolithic fuel system.

Irradiation performance of U-Mo monolithic fuel

DOE PAGES

Meyer, M. K.; Gan, J.; Jue, J. F.; ...

2014-04-01

High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. U-Mo alloys represent the best known tradeoff in these properties.more » Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.« less

Uranium-molybdenum nuclear fuel plates behaviour under heavy ion irradiation: An X-ray diffraction analysis

NASA Astrophysics Data System (ADS)

Palancher, H.; Wieschalla, N.; Martin, P.; Tucoulou, R.; Sabathier, C.; Petry, W.; Berar, J.-F.; Valot, C.; Dubois, S.

2009-03-01

Heavy ion irradiation has been proposed for discriminating UMo/Al specimens which are good candidates for research reactor fuels. Two UMo/Al dispersed fuels (U-7 wt%Mo/Al and U-10 wt%Mo/Al) have been irradiated with a 80 MeV 127I beam up to an ion fluence of 2 × 1017 cm-2. Microscopy and mainly X-ray diffraction using large and micrometer sized beams have enabled to characterize the grown interaction layer: UAl3 appears to be the only produced crystallized phase. The presence of an amorphous additional phase can however not be excluded. These results are in good agreement with characterizations performed on in-pile irradiated fuels and encourage new studies with heavy ion irradiation.

Post Irradiation TEM Investigation of ZrN Coated U(Mo) Particles Prepared with FIB

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

Van Renterghem, W.; Leenaers, A.; Van den Berghe, S.

2015-10-01

In the framework of the Selenium project, two dispersion fuel plates were fabricated with Si and ZrN coated fuel particles and irradiated in the Br2 reactor of SCK•CEN to high burn-up. The first analysis of the irradiated plate proved the reduced swelling of the fuel plate and interaction layer growth up to 70% burn-up. The question was raised how the structure of the interaction layer had been affected by the irradiation and how the structure of the fuel particles had evolved. Hereto, samples from the ZrN coated UMo particles were prepared for transmission electron microscopy (TEM) using focused ion beammore » milling (FIB) at INL. The FIB technique allowed to precisely select the area of the interaction layer and/or fuel to produce a sample that is TEM transparent over an area of 20 by 20 µm. In this contribution, the first TEM results will be presented from the 66% burn-up sample.« less

Growth kinetics and microstructural evolution during hot isostatic pressing of U-10 wt.% Mo monolithic fuel plate in AA6061 cladding with Zr diffusion barrier

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

Y. Park; J. Yoo; K. Huang

2014-04-01

Phase constituents and microstructure changes in RERTR fuel plate assemblies as functions of temperature and duration of hot-isostatic pressing (HIP) during fabrication were examined. The HIP process was carried out as functions of temperature (520, 540, 560 and 580 °C for 90 min) and time (45–345 min at 560 °C) to bond 6061 Al-alloy to the Zr diffusion barrier that had been co-rolled with U-10 wt.% Mo (U10Mo) fuel monolith prior to the HIP process. Scanning and transmission electron microscopies were employed to examine the phase constituents, microstructure and layer thickness of interaction products from interdiffusion. At the interface betweenmore » the U10Mo and Zr, following the co-rolling, the UZr2 phase was observed to develop adjacent to Zr, and the a-U phase was found between the UZr2 and U10Mo, while the Mo2Zr was found as precipitates mostly within the a-U phase. The phase constituents and thickness of the interaction layer at the U10Mo-Zr interface remained unchanged regardless of HIP processing variation. Observable growth due to HIP was only observed for the (Al,Si)3Zr phase found at the Zr/AA6061 interface, however, with a large activation energy of 457 ± 28 kJ/mole. Thus, HIP can be carried to improve the adhesion quality of fuel plate without concern for the excessive growth of the interaction layer, particularly at the U10Mo-Zr interface with the a-U, Mo2Zr, and UZr2 phases.« less

Growth kinetics and microstructural evolution during hot isostatic pressing of U-10 wt.% Mo monolithic fuel plate in AA6061 cladding with Zr diffusion barrier

NASA Astrophysics Data System (ADS)

Park, Y.; Yoo, J.; Huang, K.; Keiser, D. D.; Jue, J. F.; Rabin, B.; Moore, G.; Sohn, Y. H.

2014-04-01

Phase constituents and microstructure changes in RERTR fuel plate assemblies as functions of temperature and duration of hot-isostatic pressing (HIP) during fabrication were examined. The HIP process was carried out as functions of temperature (520, 540, 560 and 580 °C for 90 min) and time (45-345 min at 560 °C) to bond 6061 Al-alloy to the Zr diffusion barrier that had been co-rolled with U-10 wt.% Mo (U10Mo) fuel monolith prior to the HIP process. Scanning and transmission electron microscopies were employed to examine the phase constituents, microstructure and layer thickness of interaction products from interdiffusion. At the interface between the U10Mo and Zr, following the co-rolling, the UZr2 phase was observed to develop adjacent to Zr, and the α-U phase was found between the UZr2 and U10Mo, while the Mo2Zr was found as precipitates mostly within the α-U phase. The phase constituents and thickness of the interaction layer at the U10Mo-Zr interface remained unchanged regardless of HIP processing variation. Observable growth due to HIP was only observed for the (Al,Si)3Zr phase found at the Zr/AA6061 interface, however, with a large activation energy of 457 ± 28 kJ/mole. Thus, HIP can be carried to improve the adhesion quality of fuel plate without concern for the excessive growth of the interaction layer, particularly at the U10Mo-Zr interface with the α-U, Mo2Zr, and UZr2 phases.

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

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

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

2016-09-12

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

Demonstration of femtosecond laser ablation inductively coupled plasma mass spectrometry for uranium isotopic measurements in U-10Mo nuclear fuel foils

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

Havrilla, George Joseph; Gonzalez, Jhanis

2015-06-10

The use of femtosecond laser ablation inductively coupled plasma mass spectrometry was used to demonstrate the feasibility of measuring the isotopic ratio of uranium directly in U-10Mo fuel foils. The measurements were done on both the flat surface and cross sections of bare and Zr clad U-10Mo fuel foil samples. The results for the depleted uranium content measurements were less than 10% of the accepted U235/238 ratio of 0.0020. Sampling was demonstrated for line scans and elemental mapping over large areas. In addition to the U isotopic ratio measurement, the Zr thickness could be measured as well as trace elementalmore » composition if required. A number of interesting features were observed during the feasibility measurements which could provide the basis for further investigation using this methodology. The results demonstrate the feasibility of using fs-LA-ICP-MS for measuring the U isotopic ratio in U-10Mo fuel foils.« less

The effect of thermomechanical processing on second phase particle redistribution in U-10 wt%Mo

NASA Astrophysics Data System (ADS)

Hu, Xiaohua; Wang, Xiaowo; Joshi, Vineet V.; Lavender, Curt A.

2018-03-01

The multi-pass hot-rolling process of an annealed uranium-10 wt% molybdenum (U10Mo) coupon was studied by plane-strain compression finite element modeling. As-cast U10Mo typically contains second phase particles such as uranium carbides (UC) and silicides along the grain boundaries. The volume fraction of UC is typically large, while the other phases can be redissolved in the matrix by certain heat treatments. The UC particle distribution is important due to its influence on the recrystallization processes (particle stimulated nucleation) that occur during annealing between rolling passes. Unfavorable particle distribution and fracture after rolling can affect the grain size and also influence the fuel performance in the reactor. A statistical method, i.e., the two-point correlation function (2PCF), was used to analyze the carbide particle distribution after each rolling reduction. The hot rolling simulation results show that the alignment of UC particles along grain boundaries will rotate during rolling until it is parallel to the rolling direction, to form stringer-like distributions which are typically observed in rolled products that contain inclusions. 2PCF analysis shows that the interparticle spacing shrinks along the normal direction (ND) and increases along the rolling direction (RD). The simulated particle distribution is very similar to that measured experimentally for similar rolling reductions. The magnitudes of major peaks of 2PCF along the ND decrease after large reduction. The locations of major peaks indicate the inter-stringer distances. Many more small peaks appear for the 2PCF along the RD, and this is related to the neighboring particles within stringers, which are along the RD.

Analysis of intergranular fission-gas bubble-size distributions in irradiated uranium-molybdenum alloy fuel

NASA Astrophysics Data System (ADS)

Rest, J.; Hofman, G. L.; Kim, Yeon Soo

2009-04-01

An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles is used to characterize fission-gas bubble development in low-enriched U-Mo alloy fuel irradiated in the advanced test reactor in Idaho as part of the Reduced Enrichment for Research and Test Reactor (RERTR) program. Fuel burnup was limited to less than ˜7.8 at.% U in order to capture the fuel-swelling stage prior to irradiation-induced recrystallization. The model couples the calculation of the time evolution of the average intergranular bubble radius and number density to the calculation of the intergranular bubble-size distribution based on differential growth rate and sputtering coalescence processes. Recent results on TEM analysis of intragranular bubbles in U-Mo were used to set the irradiation-induced diffusivity and re-solution rate in the bubble-swelling model. Using these values, good agreement was obtained for intergranular bubble distribution compared against measured post-irradiation examination (PIE) data using grain-boundary diffusion enhancement factors of 15-125, depending on the Mo concentration. This range of enhancement factors is consistent with values obtained in the literature.

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

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

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

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

Annealing tests of in-pile irradiated oxide coated U-Mo/Al-Si dispersed nuclear fuel

NASA Astrophysics Data System (ADS)

Zweifel, T.; Valot, Ch.; Pontillon, Y.; Lamontagne, J.; Vermersch, A.; Barrallier, L.; Blay, T.; Petry, W.; Palancher, H.

2014-09-01

U-Mo/Al based nuclear fuels have been worldwide considered as a promising high density fuel for the conversion of high flux research reactors from highly enriched uranium to lower enrichment. In this paper, we present the annealing test up to 1800 °C of in-pile irradiated U-Mo/Al-Si fuel plate samples. More than 70% of the fission gases (FGs) are released during two major FG release peaks around 500 °C and 670 °C. Additional characterisations of the samples by XRD, EPMA and SEM suggest that up to 500 °C FGs are released from IDL/matrix interfaces. The second peak at 670 °C representing the main release of FGs originates from the interaction between U-Mo and matrix in the vicinity of the cladding.

Target-fueled nuclear reactor for medical isotope production

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

Coats, Richard L.; Parma, Edward J.

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

Microstructural development from interdiffusion and reaction between Usbnd Mo and AA6061 alloys annealed at 600° and 550 °C

NASA Astrophysics Data System (ADS)

Perez, E.; Keiser, D. D.; Sohn, Y. H.

2016-08-01

The U.S. Material Management and Minimization Reactor Conversion Program is developing low enrichment fuel systems encased in Al-alloy for use in research and test reactors. Monolithic fuel plates have local regions where the Usbnd Mo fuel plate may come into contact with the Al-alloy 6061 (AA6061) cladding. This results in the development of interdiffusion zones with complex microstructures with multiple phases. In this study, the microstructural development of diffusion couples, Usbnd 7 wt%Mo, Usbnd 10 wt%Mo, and Usbnd 12 wt%Mo vs. AA6061, annealed at 600 °C for 24 h and at 550 °C for 1, 5, and 20 h, were analyzed by scanning electron microscopy with x-ray energy dispersive spectroscopy. The microstructural development and kinetics were compared to diffusion couples Usbnd Mo vs. high purity Al and binary Alsbnd Si alloys. The diffusion couples developed complex interaction regions where phase development was influenced by the alloying additions of the AA6061.

Status and progress of the RERTR program in the year 2003.

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

Travelli, A.; Nuclear Engineering Division

2003-01-01

One of the most important events affecting the RERTR program during the past year was the decision by the U.S. Department of Energy to request the U.S. Congress to significantly increase RERTR program funding. This decision was prompted, at least in part, by the terrible events of September 11, 2001, and by a high-level U.S./Russian Joint Expert Group recommendation to immediately accelerate RERTR program activities in both countries, with the goal of converting all the world's research reactors to low-enriched fuel at the earliest possible time, and including both Soviet-designed and United States-designed research reactors. The U.S. Congress is expectedmore » to approve this request very soon, and the RERTR program has prepared itself well for the intense activities that the 'Accelerated RERTR Program' will require. Promising results have been obtained in the development of a fabrication process for monolithic LEU U-Mo fuel. Most existing and future research reactors could be converted to LEU with this fuel, which has a uranium density between 15.4 and 16.4 g/cm{sup 3} and yielded promising irradiation results in 2002. The most promising method hinges on producing the monolithic meat by cold-rolling a thin ingot produced by casting. The aluminum clad and the meat are bonded by friction stir welding and the cladding surface is finished by a light cold roll. This method can be applied to the production of miniplates and appears to be extendable to the production of full-size plates, possibly with intermediate anneals. Other methods planned for investigation include high temperature bonding and hot isostatic pressing. The progress achieved within the Russian RERTR program, both for the traditional tube-type elements and for the new 'universal' LEU U-Mo pin-type elements, promises to enable soon the conversion of many Russian-designed research and test reactors. Irradiation testing of both fuel types with LEU U-Mo dispersion fuels has begun. Detailed studies are in progress to define the feasibility of converting each Russian-designed research and test reactor to either fuel type. The plan for the Accelerated RERTR Program is structured to achieve LEU conversion of all HEU research reactors supplied by the United States and Russia during the next nine years. This effort will address, in addition to the fuel development and qualification, the analyses and performance/economic/safety evaluations needed to implement the conversions. In combination with this over-arching goal, the RERTR program plans to achieve at the earliest possible date qualification of LEU U-Mo dispersion fuels with uranium densities of 6 g/cm{sup 3} and 7 g/cm{sup 3}. Reactors currently using or planning to use LEU silicide fuel will rely on this fuel after termination of the FRRSNFA program, because it is acceptable to COGEMA for reprocessing. Qualification of LEU U-Mo dispersion fuels has suffered some unavoidable delays but, to accelerate it as much as possible, the RERTR program, the French CEA, and the Australian ANSTO have agreed to jointly pursue a two-element qualification test of LEU U-Mo dispersion fuel with uranium density of 7.0 g/cm{sup 3} to be performed in the Osiris reactor during 2004. The RERTR program also intends to eliminate all obstacles to the utilization of LEU in targets for isotope production, so that this important function can be performed without the need for weapons-grade materials. All of us, working together as we have for many years, can ensure that all these goals will be achieved. By promoting the efficiency and safety of research reactors while eliminating the traffic in weapons-grade uranium, we can prevent the possibility that some of this material might fall in the wrong hands. Few causes can be more deserving of our joint efforts.« less

Observations Derived From the Characterization of Monolithic Fuel Plates Irradiated as Part of the RERTR-6 Experiment

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

D. D. Keiser, Jr.; A. B. Robinson; M. R. Finlay

2007-09-01

Evaluation of the PIE results of the monolithic plates that were irradiated as part of the RERTR-6 experiment has continued. Specifically, comparisons have been made between the microstructures of the fuel plates before and after irradiation. Using the results from the rigorous characterization that was performed on the as-fabricated plates using scanning electron microscopy, it is possible to improve understanding of how monolithic fuel plates perform when they are irradiated. This paper will discuss the changes that occur, if any, in the microstructure of a monolithic fuel plate that is fabricated using techniques like what were employed for fabricating RERTR-6more » fuel plates. In addition, the performance of fuel/cladding interaction layers that were present in the fuel plates due to the fabrication process will be discussed, particularly in the context of swelling of these layers and how these layers exhibit different behaviors depending on whether the fuel alloy in the fuel plate is U-7Mo or U-10Mo.« less

Phase decomposition of γ-U (bcc) in U-10 wt% Mo fuel alloy during hot isostatic pressing of monolithic fuel plate

NASA Astrophysics Data System (ADS)

Park, Y.; Eriksson, N.; Newell, R.; Keiser, D. D.; Sohn, Y. H.

2016-11-01

Eutectoid decomposition of γ-phase (cI2) into α-phase (oC4) and γ‧-phase (tI6) during the hot isostatic pressing (HIP) of the U-10 wt% Mo (U10Mo) alloy was investigated using monolithic fuel plate samples consisting of U10Mo fuel alloy, Zr diffusion barrier and AA6061 cladding. The decomposition of the γ-phase was observed because the HIP process is carried out near the eutectoid temperature, 555 °C. Initially, a cellular structure, consisting of γ‧-phase surrounded by α-phase, developed from the destabilization of the γ-phase. The cellular structure further developed into an alternating lamellar structure of α- and γ‧-phases. Using scanning electron microscopy and transmission electron microscopy, qualitative and quantitative microstructural analyses were carried out to identify the phase constituents, and elucidate the microstructural development based on time-temperature-transformation diagram of the U10Mo alloy. The destabilization of γ -phase into α- and γ‧-phases would be minimized when HIP process was carried out with rapid ramping/cooling rate and dwell temperature higher than 560 °C.

High-energy synchrotron study of in-pile-irradiated U–Mo fuels

DOE PAGES

Miao, Yinbin; Mo, Kun; Ye, Bei; ...

2015-12-30

We report synchrotron scattering analysis results on U-7wt%Mo fuel samples irradiated in the Advanced Test Reactor to three different burnup levels. Mature fission gas bubble superlattice was observed to form at intermediate burnup. The superlattice constant was determined to be 11.7 nm and 12.1 nm by wide-angle and small-angle scattering respectively. Grain sub-division takes place throughout the irradiation and causes the collapse of the superlattice at high burnup. The bubble superlattice expands the lattice constant and acts as strong sinks of radiation induced defects. The evolution of dislocation loops was therefore suppressed until the bubble superlattice collapses.

Development and Validation of Capabilities to Measure Thermal Properties of Layered Monolithic U-Mo Alloy Plate-Type Fuel

NASA Astrophysics Data System (ADS)

Burkes, Douglas E.; Casella, Andrew M.; Buck, Edgar C.; Casella, Amanda J.; Edwards, Matthew K.; MacFarlan, Paul J.; Pool, Karl N.; Smith, Frances N.; Steen, Franciska H.

2014-07-01

The uranium-molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world's highest power research reactors from the use of high enriched uranium to low enriched uranium. One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the thermal-conductivity behavior of the fuel system as a function of temperature and expected irradiation conditions. The purpose of this paper is to verify functionality of equipment installed in hot cells for eventual measurements on irradiated uranium-molybdenum (U-Mo) monolithic fuel specimens, refine procedures to operate the equipment, and validate models to extract the desired thermal properties. The results presented here demonstrate the adequacy of the equipment, procedures, and models that have been developed for this purpose based on measurements conducted on surrogate depleted uranium-molybdenum (DU-Mo) alloy samples containing a Zr diffusion barrier and clad in aluminum alloy 6061 (AA6061). The results are in excellent agreement with thermal property data reported in the literature for similar U-Mo alloys as a function of temperature.

Irradiation performance of HTGR fuel rods in HFIR experiments HRB-7 and -8

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

Valentine, K.H.; Homan, F.J.; Long, E.L. Jr.

1977-05-01

The HRB-7 and -8 experiments were designed as a comprehensive test of mixed thorium-uranium oxide fissile particles with Th:U ratios from 0 to 8 for HTGR recycle application. In addition, fissile particles derived from Weak-Acid Resin (WAR) were tested as a potential backup type of fissile particle for HTGR recycle. These experiments were conducted at two temperatures (1250 and 1500/sup 0/C) to determine the influence of operating temperature on the performance parameters studied. The minor objectives were comparison of advanced coating designs where ZrC replaced SiC in the Triso design, testing of fuel coated in laboratory-scale equipment with fuel coatedmore » in production-scale coaters, comparison of the performance of /sup 233/U-bearing particles with that of /sup 235/U-bearing particles, comparison of the performance of Biso coatings with Triso coatings for particles containing the same type of kernel, and testing of multijunction tungsten-rhenium thermocouples. All objectives were accomplished. As a result of these experiments the mixed thorium-uranium oxide fissile kernel was replaced by a WAR-derived particle in the reference recycle design. A tentative decision to make this change had been reached before the HRB-7 and -8 capsules were examined, and the results of the examination confirmed the accuracy of the previous decision. Even maximum dilution (Th/U approximately equal to 8) of the mixed thorium-uranium oxide kernel was insufficient to prevent amoeba of the kernels at rates that are unacceptable in a large HTGR. Other results showed the performance of /sup 233/U-bearing particles to be identical to that of /sup 235/U-bearing particles, the performance of fuel coated in production-scale equipment to be at least as good as that of fuel coated in laboratory-scale coaters, the performance of ZrC coatings to be very promising, and Biso coatings to be inferior to Triso coatings relative to fission product retention.« less

IRRADIATION PERFORMANCE OF U-Mo MONOLITHIC FUEL

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

M.K. Meyer; J. Gan; J.-F. Jue

2014-04-01

High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. UMo alloys represent the best known tradeoff in these properties.more » Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.« less

Spark plasma sintering and microstructural analysis of pure and Mo doped U3Si2 pellets

NASA Astrophysics Data System (ADS)

Lopes, Denise Adorno; Benarosch, Anna; Middleburgh, Simon; Johnson, Kyle D.

2017-12-01

U3Si2 has been considered as an alternative fuel for Light Water Reactors (LWRs) within the Accident Tolerant Fuels (ATF) initiative, begun after the Fukushima-Daiichi Nuclear accidents. Its main advantages are high thermal conductivity and high heavy metal density. Despite these benefits, U3Si2 presents an anisotropic crystallographic structure and low solubility of fission products, which can result in undesirable effects under irradiation conditions. In this paper, spark plasma sintering (SPS) of U3Si2 pellets is studied, with evaluation of the resulting microstructure. Additionally, exploiting the short sintering time in SPS, a molybdenum doped pellet was produced to investigate the early stages of the Mo-U3Si2 interaction, and analyze how this fission product is accommodated in the fuel matrix. The results show that pellets of U3Si2 with high density (>95% TD) can be obtained with SPS in the temperature range of 1200°C-1300 °C. Moreover, the short time employed in this technique was found to generate a unique microstructure for this fuel, composed mainly of closed nano-pores (<1 μm) and small average grain size (∼4.5 μm). The addition of Mo (1.5 at%) demonstrated no solubility of Mo in the U3Si2 matrix. The interaction of this fission product with the fuel matrix at 1200 °C formed, in the early stages, the stoichiometric U2Mo3Si4 ternary as well as precipitates of free uranium with small quantities of dissolved Si and Mo at the front of the reaction.

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.

Particle Count Limits Recommendation for Aviation Fuel

DTIC Science & Technology

2015-10-05

Particle Counter Methodology • Particle counts are taken utilizing calibration methodologies and standardized cleanliness code ratings – ISO 11171 – ISO...Limits Receipt Vehicle Fuel Tank Fuel Injector Aviation Fuel DEF (AUST) 5695B 18/16/13 Parker 18/16/13 14/10/7 Pamas / Parker / Particle Solutions 19/17...12 U.S. DOD 19/17/14/13* Diesel Fuel World Wide Fuel Charter 5th 18/16/13 DEF (AUST) 5695B 18/16/13 Caterpillar 18/16/13 Detroit Diesel 18/16/13 MTU

Microstructural development from interdiffusion and reaction between U–Mo and AA6061 alloys annealed at 600° and 550 °C

DOE PAGES

Perez, E.; Keiser, D. D.; Sohn, Y. H.

2016-05-10

The U.S. Material Management and Minimization Reactor Conversion Program is developing low enrichment fuel systems encased in Al-alloy for use in research and test reactors. Monolithic fuel plates have local regions where the Usingle bondMo fuel plate may come into contact with the Al-alloy 6061 (AA6061) cladding. This results in the development of interdiffusion zones with complex microstructures with multiple phases. In this study, the microstructural development of diffusion couples, U–7 wt%Mo, U–10 wt%Mo, and U–12 wt%Mo vs. AA6061, annealed at 600 °C for 24 h and at 550 °C for 1, 5, and 20 h, were analyzed by scanningmore » electron microscopy with x-ray energy dispersive spectroscopy. The microstructural development and kinetics were compared to diffusion couples U–Mo vs. high purity Al and binary Al–Si alloys. As a result, the diffusion couples developed complex interaction regions where phase development was influenced by the alloying additions of the AA6061.« less

Evaluation of Uranium-235 Measurement Techniques

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

Kaspar, Tiffany C.; Lavender, Curt A.; Dibert, Mark W.

2017-05-23

Monolithic U-Mo fuel plates are rolled to final fuel element form from the original cast ingot, and thus any inhomogeneities in 235U distribution present in the cast ingot are maintained, and potentially exaggerated, in the final fuel foil. The tolerance for inhomogeneities in the 235U concentration in the final fuel element foil is very low. A near-real-time, nondestructive technique to evaluate the 235U distribution in the cast ingot is required in order to provide feedback to the casting process. Based on the technical analysis herein, gamma spectroscopy has been recommended to provide a near-real-time measure of the 235U distribution inmore » U-Mo cast plates.« 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

Interdiffusion in U 3Si-Al, U 3Si 2-Al, and USi-Al dispersion fuels during irradiation

NASA Astrophysics Data System (ADS)

Kim, Yeon Soo; Hofman, Gerard L.

2011-03-01

Uranium-silicide compound fuel dispersion in an Al matrix is used in research and test reactors worldwide. Interaction layer (IL) growth between fuel particles and the matrix is one of performance issues. The interaction layer growth data for U 3Si, U 3Si 2 and USi dispersions in Al were obtained from both out-of-pile and in-pile tests. The IL is dominantly U(AlSi) 3 from out-of-pile tests, but its (Al + Si)/U ratio from in-pile tests is higher than the out-of-pile data, because of amorphous behavior of the ILs. IL growth correlations were developed for U 3Si-Al and U 3Si 2-Al. The IL growth rates were dependent on the U/Si ratio of the fuel compounds. During irradiation, however, the IL growth rates did not decrease with the decreasing U/Si ratio by fission. It is reasoned that transition metal fission products in the IL compensate the loss of U atoms by providing chemical potential for Al diffusion and volume expansion by solid swelling and gas bubble swelling. The addition of Mo in U 3Si 2 reduces the IL growth rate, which is similar to that of UMo alloy dispersion in a silicon-added Al matrix.

Development of Nitride Coating Using Atomic Layer Deposition for Low-Enriched Uranium Fuel Powder

NASA Astrophysics Data System (ADS)

Bhattacharya, Sumit

High-performance research reactors require fuel that operates at high specific power and can withstand high fission density, but at relatively low temperatures. The design of the research reactor fuels is done for efficient heat emission, and consists of assemblies of thin-plates cladding made from aluminum alloy. The low-enriched fuels (LEU) were developed for replacing high-enriched fuels (HEU) for these reactors necessitates a significantly increased uranium density in the fuel to counterbalance the decrease in enrichment. One of the most promising new fuel candidate is U-Mo alloy, in a U-Mo/Al dispersion fuel form, due to its high uranium loading as well as excellent irradiation resistance performance, is being developed extensively to convert from HEU fuel to LEU fuel for high-performance research reactors. However, the formation of an interaction layer (IL) between U-Mo particles and the Al matrix, and the associated pore formation, under high heat flux and high burnup conditions, degrade the irradiation performance of the U-Mo/Al dispersion fuel. From the recent tests results accumulated from the surface engineering of low enriched uranium fuel (SELENIUM) and MIR reactor displayed that a surface barrier coating like physical vapor deposited (PVD) zirconium nitride (ZrN) can significantly reduce the interaction layer. The barrier coating performed well at low burn up but above a fluence rate of 5x 1021 ions/cm2 the swelling reappeared due to formation interaction layer. With this result in mind the objective of this research was to develop an ultrathin ZrN coating over particulate uranium-molybdenum nuclear fuel using a modified savannah 200 atomic layer deposition (ALD) system. This is done in support of the US Department of Energy's (DOE) effort to slow down the interaction at fluence rate and reach higher burn up for high power research reactor. The low-pressure Savannah 200 ALD system is modified to be designed as a batch powder coating system using the metal organic chemical precursors tetrakis dimethylamido zirconium (TDMAZr) and ammonia( NH3) for succesful deposition of ZrN coating. Nitrogen (N2) gas carried the chemicals to a hot wall reactor maintained at a temperature range of 235 to 245 °C. The ALD system design evolved over the course of this research as the process variables were steadily improved. The conditions found deemed for attaining best coating were at a temperature of 245 °C, with pulse time of 0.8 seconds for TDMAZr and 0.1 seconds for NH3 along with 15 seconds of purge time in-between each cycle. The ALD system was successful in making 1-micrometer (um) ZrN with low levels of chemical impurities over U-Mo powder batches. The deposited coatings were characterized using scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS), electron energy loss spectroscopy (EELS) and Transmission electron microscope (TEM). This document describes the establishment of the Savannah 200 ALD system, precursor surface reaction procedures and finally the nature of the coating achieved, including characterization of the coating at the different stages of deposition. It was found that an interlayer of alumina in between ZrN and the U-Mo surface was required to reduce the residual stress generated during the ALD procedure. The alumina not only removed the risk of cracking and spallation of the ZrN coating but also provided adequate strength for the barrier layer to withstand the fuel plate rolling conditions. The ZrN coating was nano crystalline in nature, with grain size varying from 5-10 nm, the deposited layer was found to be dense consisting of a layered structure. The coating could retain its crystallinity and maintain its phase when irradiated with 1 MeV single charged ion Kr to produce a damage of 10 displacement per atom (DPA) at intermediate voltage electron microscopy (IVEM).

Progress of the RERTR program in 2001.

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

Travelli, A.

2002-03-07

This paper describes the 2001 progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners. Postirradiation examinations of microplates have continued to reveal excellent irradiation behavior of U-Mo dispersion fuels in a variety of compositions and irradiating conditions. Irradiation of two new batches of miniplates of greater sizes was completed in the ATR to investigate the swelling behavior of these fuels under prototypic conditions. These materials hold the promise of achieving the program goal of developing LEU research reactor fuels with uranium densities in the 8-9 g/cm{sup 3} range. Qualificationmore » of the U-Mo dispersion fuels has been delayed by a patent issue involving KAERI. Test fuel elements with uranium density of 6 g/cm{sup 3} are being fabricated by BWXT and are expected to begin undergoing irradiation in the HFR-Petten reactor around March 2003, with a goal of qualifying this fuel by mid-2005. U-Mo fuel with uranium density of 8-9 g/cm{sup 3} is expected to be qualified by mid-2007. Final irradiation tests of LEU {sup 99}Mo targets in the RAS-GAS reactor at BATAN, in Indonesia, had to be postponed because of the 9/11 attacks, but the results collected to date indicate that these targets will soon be ready for commercial production. Excellent cooperation is also in progress with the CNEA in Argentina, MDSN/AECL in Canada, and ANSTO in Australia. Irradiation testing of five WWR-M2 tube-type fuel assemblies fabricated by the NZChK and containing LEU UO{sub 2} dispersion fuel was successfully completed within the Russian RERTR program. A new LEU U-Mo pin-type fuel that could be used to convert most Russian-designed research reactors has been developed by VNIINM and is ready for testing. Four additional shipments containing 822 spent fuel assemblies from foreign research reactors were accepted by the U.S. by September 30, 2001. Altogether, 4,562 spent fuel assemblies from foreign research reactors had been received by that date by the U.S. under the FRR SNF acceptance policy. The RERTR program is aggressively pursuing qualification of high-density LEU U-Mo dispersion fuels, with the dual goal of enabling further conversions and of developing a substitute for LEU silicide fuels that can be more easily disposed of after expiration of the U.S. FRR SNF Acceptance Program. As in the past, the success of the RERTR program will depend on the international friendship and cooperation that has always been its trademark.« less

Neutron diffraction measurement of residual stresses, dislocation density and texture in Zr-bonded U-10Mo “mini” fuel foils and plates

DOE PAGES

Brown, Donald William; Okuniewski, Maria A.; Sisneros, Thomas A.; ...

2016-12-01

Here, Al clad U-10Mo fuel plates are being considered for conversion of several research reactors from high-enriched to low-enriched U fuel. Neutron diffraction measurements of the textures, residual phase stresses, and dislocation densities in the individual phases of the mini-foils throughout several processing steps and following hot-isostatic pressing to the Al cladding, have been completed. Recovery and recrystallization of the bare U-10Mo fuel foil, as indicated by the dislocation density and texture, are observed depending on the state of the material prior to annealing and the duration and temperature of the annealing process. In general, the cladding procedure significantly reducesmore » the dislocation density, but the final state of the clad plate, both texture and dislocation density, depends strongly on the final processing step of the fuel foil. In contrast, the residual stress state of the final plate is dominated by the thermal expansion mismatch of the constituent materials.« less

Neutron diffraction measurement of residual stresses, dislocation density and texture in Zr-bonded U-10Mo “mini” fuel foils and plates

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

Brown, Donald William; Okuniewski, Maria A.; Sisneros, Thomas A.

Here, Al clad U-10Mo fuel plates are being considered for conversion of several research reactors from high-enriched to low-enriched U fuel. Neutron diffraction measurements of the textures, residual phase stresses, and dislocation densities in the individual phases of the mini-foils throughout several processing steps and following hot-isostatic pressing to the Al cladding, have been completed. Recovery and recrystallization of the bare U-10Mo fuel foil, as indicated by the dislocation density and texture, are observed depending on the state of the material prior to annealing and the duration and temperature of the annealing process. In general, the cladding procedure significantly reducesmore » the dislocation density, but the final state of the clad plate, both texture and dislocation density, depends strongly on the final processing step of the fuel foil. In contrast, the residual stress state of the final plate is dominated by the thermal expansion mismatch of the constituent materials.« less

Microstructure of RERTR DU-alloys irradiated with krypton ions up to 100 dpa

NASA Astrophysics Data System (ADS)

Gan, J.; Keiser, D. D., Jr.; Miller, B. D.; Wachs, D. M.; Allen, T. R.; Kirk, M.; Rest, J.

2011-04-01

The radiation stability of the interaction product formed at the fuel-matrix interface of research reactor dispersion fuels, under fission-product bombardment, has a strong impact on fuel performance. Three depleted uranium alloys were cast that consisted of the following five phases to be investigated: U(Si, Al) 3, (U, Mo)(Si, Al) 3, UMo 2Al 20, U 6Mo 4Al 43, and UAl 4. Irradiation of transmission electron microscopy (TEM) disc samples with 500-keV Kr ions at 200 °C to doses up to ˜100 displacements per atom (dpa) were conducted using a 300-keV electron microscope equipped with an ion accelerator. TEM results show that the U(Si, Al) 3 and UAl 4 phases remain crystalline at 100 dpa without forming voids. The (U, Mo)(Si, Al) 3 and UMo 2Al 20 phases become amorphous at 1 and ˜2 dpa, respectively, and show no evidence of voids at 100 dpa. The U 6Mo 4Al 43 phase goes to amorphous at less than 1 dpa and reveals high density voids at 100 dpa.

Mesoscale model for fission-induced recrystallization in U-7Mo alloy

DOE PAGES

Liang, Linyun; Mei, Zhi -Gang; Kim, Yeon Soo; ...

2016-08-09

A mesoscale model is developed by integrating the rate theory and phase-field models and is used to study the fission-induced recrystallization in U-7Mo alloy. The rate theory model is used to predict the dislocation density and the recrystallization nuclei density due to irradiation. The predicted fission rate and temperature dependences of the dislocation density are in good agreement with experimental measurements. This information is used as input for the multiphase phase-field model to investigate the fission-induced recrystallization kinetics. The simulated recrystallization volume fraction and bubble induced swelling agree well with experimental data. The effects of the fission rate, initial grainmore » size, and grain morphology on the recrystallization kinetics are discussed based on an analysis of recrystallization growth rate using the modified Avrami equation. Here, we conclude that the initial microstructure of the U-Mo fuels, especially the grain size, can be used to effectively control the rate of fission-induced recrystallization and therefore swelling.« less

Physical properties of monolithic U8 wt.%-Mo

NASA Astrophysics Data System (ADS)

Hengstler, R. M.; Beck, L.; Breitkreutz, H.; Jarousse, C.; Jungwirth, R.; Petry, W.; Schmid, W.; Schneider, J.; Wieschalla, N.

2010-07-01

As a possible high density fuel for research reactors, monolithic U8 wt.%-Mo ("U8Mo") was examined with regard to its structural, thermal and electric properties. X-ray diffraction by the Bragg-Brentano method was used to reveal the tetragonal lattice structure of rolled U8Mo. The specific heat capacity of cast U8Mo was determined by differential scanning calorimetry, its thermal diffusivity was measured by the laser flash method and its mass density by Archimedes' principle. From these results, the thermal conductivity of U8Mo in the temperature range from 40 °C to 250 °C was calculated; in the measured temperature range, it is in good accordance with literature data for UMo with 8 and 9 wt.% Mo, is higher than for 10 wt.% Mo and lower than for 5 wt.% Mo. The electric conductivity of rolled and cast U8Mo was measured by a four-wire method and the electron based part of the thermal conductivity calculated by the Wiedemann-Frantz law. Rolled and cast U8Mo was irradiated at about 150 °C with 80 MeV 127I ions to receive the same iodine ion density in the damage peak region as the fission product density in the fuel of a typical high flux reactor after the targeted nuclear burn-up. XRD analysis of irradiated U8Mo showed a change of the lattice parameters as well as the creation of UO 2 in the superficial sample regions; however, a phase change by irradiation was not observed. The determination of the electron based part of the thermal conductivity of the irradiated samples failed due to high measurement errors which are caused by the low thickness of the damage region in the ion irradiated samples.

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

Stillman, J. A.; Feldman, E. E.; Wilson, E. H.

This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. This report contains themore » results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. In the framework of non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context most research and test reactors, both domestic and international, have started a program of conversion to the use of LEU fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (U-Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MURR. This report presents the results of a study of core behavior under a set of accident conditions for MURR cores fueled with HEU U-Alx dispersion fuel or LEU monolithic U-Mo alloy fuel with 10 wt% Mo (U-10Mo).« less

ZPR-6 assembly 7 high {sup 240}Pu core experiments : a fast reactor core with mixed (Pu,U)-oxide fuel and a centeral high{sup 240}Pu zone.

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

Lell, R. M.; Morman, J. A.; Schaefer, R.W.

ZPR-6 Assembly 7 (ZPR-6/7) encompasses a series of experiments performed at the ZPR-6 facility at Argonne National Laboratory in 1970 and 1971 as part of the Demonstration Reactor Benchmark Program (Reference 1). Assembly 7 simulated a large sodium-cooled LMFBR with mixed oxide fuel, depleted uranium radial and axial blankets, and a core H/D near unity. ZPR-6/7 was designed to test fast reactor physics data and methods, so configurations in the Assembly 7 program were as simple as possible in terms of geometry and composition. ZPR-6/7 had a very uniform core assembled from small plates of depleted uranium, sodium, iron oxide,more » U{sub 3}O{sub 8} and Pu-U-Mo alloy loaded into stainless steel drawers. The steel drawers were placed in square stainless steel tubes in the two halves of a split table machine. ZPR-6/7 had a simple, symmetric core unit cell whose neutronic characteristics were dominated by plutonium and {sup 238}U. The core was surrounded by thick radial and axial regions of depleted uranium to simulate radial and axial blankets and to isolate the core from the surrounding room. The ZPR-6/7 program encompassed 139 separate core loadings which include the initial approach to critical and all subsequent core loading changes required to perform specific experiments and measurements. In this context a loading refers to a particular configuration of fueled drawers, radial blanket drawers and experimental equipment (if present) in the matrix of steel tubes. Two principal core configurations were established. The uniform core (Loadings 1-84) had a relatively uniform core composition. The high {sup 240}Pu core (Loadings 85-139) was a variant on the uniform core. The plutonium in the Pu-U-Mo fuel plates in the uniform core contains 11% {sup 240}Pu. In the high {sup 240}Pu core, all Pu-U-Mo plates in the inner core region (central 61 matrix locations per half of the split table machine) were replaced by Pu-U-Mo plates containing 27% {sup 240}Pu in the plutonium component to construct a central core zone with a composition closer to that in an LMFBR core with high burnup. The high {sup 240}Pu configuration was constructed for two reasons. First, the composition of the high {sup 240}Pu zone more closely matched the composition of LMFBR cores anticipated in design work in 1970. Second, comparison of measurements in the ZPR-6/7 uniform core with corresponding measurements in the high {sup 240}Pu zone provided an assessment of some of the effects of long-term {sup 240}Pu buildup in LMFBR cores. The uniform core version of ZPR-6/7 is evaluated in ZPR-LMFR-EXP-001. This document only addresses measurements in the high {sup 240}Pu core version of ZPR-6/7. Many types of measurements were performed as part of the ZPR-6/7 program. Measurements of criticality, sodium void worth, control rod worth and reaction rate distributions in the high {sup 240}Pu core configuration are evaluated here. For each category of measurements, the uncertainties are evaluated, and benchmark model data are provided.« less

U-Mo Monolithic Fuel for Nuclear Research and Test Reactors

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

Prabhakaran, Ramprashad

The metallic fuel selected to replace the current HEU fuels in the research and test reactors is the LEU-10 weight % Mo alloy in the form of a thin sheet or foil encapsulated in AA6061 aluminum alloy with a zirconium interlayer. In order to effectively lead this pursuit, new developments in processing and fabrication of the fuel elements have been initiated, along with a better understanding of material behavior before and after irradiation as a result of these new developments. This editorial note gives an introduction about research and test reactors, need for HEU to LEU conversion, fuel requirements, highmore » uranium density monolithic fuel development and an overview of the four articles published in the December 2017 issue of JOM under a special topic titled “U-Mo Monolithic Fuel for Nuclear Research and Test Reactors”.« less

Development of PRIME for irradiation performance analysis of U-Mo/Al dispersion fuel

NASA Astrophysics Data System (ADS)

Jeong, Gwan Yoon; Kim, Yeon Soo; Jeong, Yong Jin; Park, Jong Man; Sohn, Dong-Seong

2018-04-01

A prediction code for the thermo-mechanical performance of research reactor fuel (PRIME) has been developed with the implementation of developed models to analyze the irradiation behavior of U-Mo dispersion fuel. The code is capable of predicting the two-dimensional thermal and mechanical performance of U-Mo dispersion fuel during irradiation. A finite element method was employed to solve the governing equations for thermal and mechanical equilibria. Temperature- and burnup-dependent material properties of the fuel meat constituents and cladding were used. The numerical solution schemes in PRIME were verified by benchmarking solutions obtained using a commercial finite element analysis program (ABAQUS). The code was validated using irradiation data from RERTR, HAMP-1, and E-FUTURE tests. The measured irradiation data used in the validation were IL thickness, volume fractions of fuel meat constituents for the thermal analysis, and profiles of the plate thickness changes and fuel meat swelling for the mechanical analysis. The prediction results were in good agreement with the measurement data for both thermal and mechanical analyses, confirming the validity of the code.

Status and progress of the RERTR program in the year 2000.

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

Travelli, A.

2000-09-28

This paper describes the progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners during the year 2000 and discusses the main activities planned for the year 2001. The past year was characterized by important accomplishments and events for the RERTR program. Four additional shipments containing 503 spent fuel assemblies from foreign research reactors were accepted by the U.S. Altogether, 3,740 spent fuel assemblies from foreign research reactors have been received by the U.S. under the acceptance policy. Postirradiation examinations of three batches of microplates have continued to reveal excellentmore » irradiation behavior of U-MO dispersion fuels in a variety of compositions and irradiating conditions. h-radiation of two new batches of miniplates of greater sizes is in progress in the ATR to investigate me swelling behavior of these fuels under prototypic conditions. These materials hold the promise of achieving the program goal of developing LEU research reactor fuels with uranium densities in the 8-9 g /cm{sup 3} range. Qualification of the U-MO dispersion fuels is proceeding on schedule. Test fuel elements with 6 gU/cm{sup 3} are being fabricated by BWXT and are scheduled to begin undergoing irradiation in the HFR-Petten in the spring of 2001, with a goal of qualifying this fuel by the end of 2003. U-Mo with 8-9 gU/cm{sup 3} is planned to be qualified by the end of 2005. Joint LEU conversion feasibility studies were completed for HFR-Petten and for SAFARI-1. Significant improvements were made in the design of LEU metal-foil annular targets that would allow efficient production of fission {sup 99}Mo. Irradiations in the RAS-GAS reactor showed that these targets can formed from aluminum tubes, and that the yield and purity of their product from the acidic process were at least as good as those from the HEU Cintichem targets. Progress was made on irradiation testing of LEU UO{sub 2} dispersion fuel and on LEU conversion feasibility studies in the Russian RERTR program. Conversion of the BER-11reactor in Berlin, Germany, was completed and conversion of the La Reins reactor in Santiago, Chile, began. These are exciting times for the program. In the fuel development area, the RERTR program is aggressively pursuing qualification of high-density LEU U-Mo dispersion fuels, with the dual goal of enabling fi.uther conversions and of developing a substitute for LEU silicide fuels that can be more easily disposed of after expiration of the FRR SNF Acceptance Program. The {sup 99}Mo effort has reached the point where it appears feasible for all the {sup 99}Mo producers of the world to agree jointly to a common course of action leading to the elimination of HEU use in their processes. As in the past, the success of the RERTR program will depend on the international friendship and cooperation that has always been its trademark.« less

MTR plates modeling with MAIA

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

Marelle, V.; Dubois, S.; Ripert, M.

2008-07-15

MAIA is a thermo-mechanical code dedicated to the modeling of MTR fuel plates. The main physical phenomena modeled in the code are the cladding oxidation, the interaction between fuel and Al-matrix, the swelling due to fission products and the Al/fuel particles interaction. The creeping of the plate can be modeled in the mechanical calculation. MAIA has been validated on U-Mo dispersion fuel experiments such as IRIS 1 and 2 and FUTURE. The results are in rather good agreement with post-irradiation examinations. MAIA can also be used to calculate in-pile behavior of U{sub 3}Si{sub 2} plates as in the SHARE experimentmore » irradiated in the SCK/Mol BR2 reactor. The main outputs given by MAIA throughout the irradiation are temperatures, cladding oxidation thickness, interaction thickness, volume fraction of meat constituents, swelling, displacements, strains and stresses. MAIA is originally a two-dimensional code but a three-dimensional version is currently under development. (author)« less

Advantages and Disadvantages of using a Focused Ion Beam to Prepare TEM Samples From Irradiated U-10Mo Monolithic Nuclear Fuel

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

B. D. Miller; J. Gan; J. Madden

2012-05-01

Transmission electron microscopy (TEM), scanning electron microscopy (SEM), and focused ion beam (FIB) milling were performed on an irradiated U-10Mo monolithic fuel to understand its irradiation microstructure. This is the first reported TEM work of irradiated fuel sample prepared using a FIB. Advantages and disadvantages of using the FIB to create TEM samples from this irradiated fuel will be presented along with some results from the work. Sample preparation techniques used to create SEM and FIB samples from the brittle irradiated monolithic sample will also be discussed.

Development of PRIME for irradiation performance analysis of U-Mo/Al dispersion fuel

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

Jeong, Gwan Yoon; Kim, Yeon Soo; Jeong, Yong Jin

A prediction code for the thermo-mechanical performance of research reactor fuel (PRIME) has been developed with the implementation of developed models to analyze the irradiation behavior of U-Mo dispersion fuel. The code is capable of predicting the two-dimensional thermal and mechanical performance of U-Mo dispersion fuel during irradiation. A finite element method was employed to solve the governing equations for thermal and mechanical equilibria. Temperature-and burnup-dependent material properties of the fuel meat constituents and cladding were used. The numerical solution schemes in PRIME were verified by benchmarking solutions obtained using a commercial finite element analysis program (ABAQUS).The code was validatedmore » using irradiation data from RERTR, HAMP-1, and E-FUTURE tests. The measured irradiation data used in the validation were IL thickness, volume fractions of fuel meat constituents for the thermal analysis, and profiles of the plate thickness changes and fuel meat swelling for the mechanical analysis. The prediction results were in good agreement with the measurement data for both thermal and mechanical analyses, confirming the validity of the code. (c) 2018 Elsevier B.V. All rights reserved.« less

Post-Irradiation Non-Destructive Analyses of the AFIP-7 Experiment

NASA Astrophysics Data System (ADS)

Williams, W. J.; Robinson, A. B.; Rabin, B. H.

2017-12-01

This article reports the results and interpretation of post-irradiation non-destructive examinations performed on four curved full-size fuel plates that comprise the AFIP-7 experiment. These fuel plates, having a U-10 wt.%Mo monolithic design, were irradiated under moderate operating conditions in the Advanced Test Reactor to assess fuel performance for geometries that are prototypic of research reactor fuel assemblies. Non-destructive examinations include visual examination, neutron radiography, profilometry, and precision gamma scanning. This article evaluates the qualitative and quantitative data taken for each plate, compares corresponding data sets, and presents the results of swelling analyses. These characterization results demonstrate that the fuel meets established irradiation performance requirements for mechanical integrity, geometric stability, and stable and predictable behavior.

Modeling the Homogenization Kinetics of As-Cast U-10wt% Mo alloys

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

Xu, Zhijie; Joshi, Vineet; Hu, Shenyang Y.

2016-01-15

Low-enriched U-22at% Mo (U-10Mo) alloy has been considered as an alternative material to replace the highly enriched fuels in research reactors. For the U-10Mo to work effectively and replace the existing fuel material, a thorough understanding of the microstructure development from as-cast to the final formed structure is required. The as-cast microstructure typically resembles an inhomogeneous microstructure with regions containing molybdenum-rich and -lean regions, which may affect the processing and possibly the in-reactor performance. This as-cast structure must be homogenized by thermal treatment to produce a uniform Mo distribution. The development of a modeling capability will improve the understanding ofmore » the effect of initial microstructures on the Mo homogenization kinetics. In the current work, we investigated the effect of as-cast microstructure on the homogenization kinetics. The kinetics of the homogenization was modeled based on a rigorous algorithm that relates the line scan data of Mo concentration to the gray scale in energy dispersive spectroscopy images, which was used to generate a reconstructed Mo concentration map. The map was then used as realistic microstructure input for physics-based homogenization models, where the entire homogenization kinetics can be simulated and validated against the available experiment data at different homogenization times and temperatures.« less

Neutronics Conversion Analyses of the Laue-Langevin Institute (ILL) High Flux Reactor (RHF)

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

Bergeron, A.; Dionne, B.; Calzavara, Y.

2014-09-30

The following report describes the neutronics results obtained with the MCNP model of the RHF U7Mo LEU reference design that has been established in 2010 during the feasibility analysis. This work constitutes a complete and detailed neutronics analysis of that LEU design using models that have been significantly improved since 2010 and the release of the feasibility report. When possible, the credibility of the neutronics model is tested by comparing the HEU model results with experimental data or other codes calculations results. The results obtained with the LEU model are systematically compared to the HEU model. The changes applied tomore » the neutronics model lead to better comparisons with experimental data or improved the calculation efficiency but do not challenge the conclusion of the feasibility analysis. If the U7Mo fuel is commercially available, not cost prohibitive, a back-end solution is established and if it is possible to manufacture the proposed element, neutronics analyses show that the performance of the reactor would not be challenged by the conversion to LEU fuel.« less

Army Demonstration of Light Obscuration Particle Counters for Monitoring Aviation Fuel Contamination

DTIC Science & Technology

2013-05-07

Hydraulic industry has utilized this technology for decades and created a mature process •Hydraulic industry has developed recognized calibration ...Vehicle Fuel Tank Fuel Injector Aviation Fuel DEF (AUST) 5695B 18/16/13 Parker 18/16/13 14/10/7 Pamas/Parker/Particle Solutions 19/17/12 U.S. Army 19...17/14/13* Diesel Fuel World Wide Fuel Charter 4th 18/16/13 DEF (AUST) 5695B 18/16/13 Bosch/Cummins 18/16/13 Donaldson 22/21/18 14/13/11 12/9/6 P ll

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

Burkes, Douglas E.; Senor, David J.; Casella, Andrew M.

Numerous global programs are focused on the continued development of existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world’s remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. Some of these programs are focused on development and qualification of a fuel design that consists of a uranium-molybdenum (U-Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. The current paper extends a failure model originally developed for UO2-stainless steel dispersion fuels and used currently available thermal-mechanical property information for the materials ofmore » interest in the current proposed design. A number of fabrication and irradiation parameters were investigated to understand the conditions at which failure of the matrix, classified as pore formation in the matrix, might occur. The results compared well with experimental observations published as part of the Reduced Enrichment for Research and Test Reactors (RERTR)-6 and -7 mini-plate experiments. Fission rate, a function of the 235U enrichment, appeared to be the most influential parameter in premature failure, mainly as a result of increased interaction layer formation and operational temperature, which coincidentally decreased the yield strength of the matrix and caused more rapid fission gas production and recoil into the surrounding matrix material. Addition of silicon to the matrix appeared effective at reducing the rate of interaction layer formation and can extend the performance of a fuel plate under a certain set of irradiation conditions, primarily moderate heat flux and burnup. Increasing the dispersed fuel particle diameter may also be effective, but only when combined with other parameters, e.g., lower enrichment and increased Si concentration. The model may serve as a valuable tool in initial experimental design.« less

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

Devaraj, Arun; Jana, Saumyadeep; McInnis, Colleen A.

During eutectoid transformation of U-10Mo alloy, uniform metastable γ UMo phase is expected to transform to a mixture of α-U and γ’-U 2Mo phase. The presence of transformation products in final U-10Mo fuel, especially the α phase is considered detrimental for fuel irradiation performance, so it is critical to accurately evaluate the extent of transformation in the final U-10Mo alloy. This phase transformation can cause a volume change that induces a density change in final alloy. To understand this density and volume change, we developed a theoretical model to calculate the volume expansion and resultant density change of U-10Mo alloymore » as a function of the extent of eutectoid transformation. Based on the theoretically calculated density change for 0 to 100% transformation, we conclude that an experimental density measurement system will be challenging to employ to reliably detect and quantify the extent of transformation. Subsequently, to assess the ability of various methods to detect the transformation in U-10Mo, we annealed U-10Mo alloy samples at 500°C for various times to achieve in low, medium, and high extent of transformation. After the heat treatment at 500°C, the samples were metallographically polished and subjected to optical microscopy and x-ray diffraction (XRD) methods. Based on our assessment, optical microscopy and image processing can be used to determine the transformed area fraction, which can then be correlated with the α phase volume fraction measured by XRD analysis. XRD analysis of U-10Mo aged at 500°C detected only α phase and no γ’ was detected. To further validate the XRD results, atom probe tomography (APT) was used to understand the composition of transformed regions in U-10Mo alloys aged at 500°C for 10 hours. Based on the APT results, the lamellar transformation product was found to comprise α phase with close to 0 at% Mo and γ phase with 28–32 at% Mo, and the Mo concentration was highest at the α/γ interface.« less

New generation nuclear fuel structures: Dense particles in selectively soluble matrix

NASA Astrophysics Data System (ADS)

Devlin, Dave; Jarvinen, Gordon; Patterson, Brian; Pattillo, Steve; Valdez, James; Liu, X.-Y.; Phillips, Jonathan

2009-11-01

We have developed a technology for dispersing sub-millimeter sized fuel particles within a bulk matrix that can be selectively dissolved. This may enable the generation of advanced nuclear fuels with easy separation of actinides and fission products. The large kinetic energy of the fission products results in most of them escaping from the sub-millimeter sized fuel particles and depositing in the matrix during burning of the fuel in the reactor. After the fuel is used and allowed to cool for a period of time, the matrix can be dissolved and the fission products removed for disposal while the fuel particles are collected by filtration for recycle. The success of such an approach would meet a major goal of the GNEP program to provide advanced recycle technology for nuclear energy production. The benefits of such an approach include (1) greatly reduced cost of the actinide/fission product separation process, (2) ease of recycle of the fuel particles, and (3) a radiation barrier to prevent theft or diversion of the recycled fuel particles during the time they are re-fabricated into new fuel. In this study we describe a method to make surrogate nuclear fuels of micrometer scale W (shell)/Mo (core) or HfO 2 particles embedded in an MgO matrix that allows easy separation of the fission products and their embedded particles. In brief, the method consists of physically mixing W-Mo or hafnia particles with an MgO precursor. Heating the mixture, in air or argon, without agitation, to a temperature is required for complete decomposition of the precursor. The resulting material was examined using chemical analysis, scanning electron microscopy, X-ray diffraction and micro X-ray computed tomography and found to consist of evenly dispersed particles in an MgO + matrix. We believe this methodology can be extended to actinides and other matrix materials.

Combinatorial discovery of new methanol-tolerant non-noble metal cathode electrocatalysts for direct methanol fuel cells.

PubMed

Yu, Jong-Sung; Kim, Min-Sik; Kim, Jung Ho

2010-12-14

Combinatorial synthesis and screening were used to identify methanol-tolerant non-platinum cathode electrocatalysts for use in direct methanol fuel cells (DMFCs). Oxygen reduction consumes protons at the surface of DMFC cathode catalysts. In combinatorial screening, this pH change allows one to differentiate active catalysts using fluorescent acid-base indicators. Combinatorial libraries of carbon-supported catalyst compositions containing Ru, Mo, W, Sn, and Se were screened. Ternary and quaternary compositions containing Ru, Sn, Mo, Se were more active than the "standard" Alonso-Vante catalyst, Ru(3)Mo(0.08)Se(2), when tested in liquid-feed DMFCs. Physical characterization of the most active catalysts by powder X-ray diffraction, gas adsorption, and X-ray photoelectron spectroscopy revealed that the predominant crystalline phase was hexagonal close-packed (hcp) ruthenium, and showed a surface mostly covered with oxide. The best new catalyst, Ru(7.0)Sn(1.0)Se(1.0), was significantly more active than Ru(3)Se(2)Mo(0.08), even though the latter contained smaller particles.

Effect of Silicon in U-10Mo Alloy

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

Kautz, Elizabeth J.; Devaraj, Arun; Kovarik, Libor

2017-08-31

This document details a method for evaluating the effect of silicon impurity content on U-10Mo alloys. Silicon concentration in U-10Mo alloys has been shown to impact the following: volume fraction of precipitate phases, effective density of the final alloy, and 235-U enrichment in the gamma-UMo matrix. This report presents a model for calculating these quantities as a function of Silicon concentration, which along with fuel foil characterization data, will serve as a reference for quality control of the U-10Mo final alloy Si content. Additionally, detailed characterization using scanning electron microscope imaging, transmission electron microscope diffraction, and atom probe tomography showedmore » that Silicon impurities present in U-10Mo alloys form a Si-rich precipitate phase.« less

Characterization of fission gas bubbles in irradiated U-10Mo fuel

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

Casella, Andrew M.; Burkes, Douglas E.; MacFarlan, Paul J.

2017-09-01

Irradiated U-10Mo fuel samples were prepared with traditional mechanical potting and polishing methods with in a hot cell. They were then removed and imaged with an SEM located outside of a hot cell. The images were then processed with basic imaging techniques from 3 separate software packages. The results were compared and a baseline method for characterization of fission gas bubbles in the samples is proposed. It is hoped that through adoption of or comparison to this baseline method that sample characterization can be somewhat standardized across the field of post irradiated examination of metal fuels.

Observed Changes in As-Fabricated U-10Mo Monolithic Fuel Microstructures After Irradiation in the Advanced Test Reactor

NASA Astrophysics Data System (ADS)

Keiser, Dennis; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Madden, James

2017-12-01

A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Management and Minimization Program, earlier known as the Reduced Enrichment for Research and Test Reactors Program, for utilization in research and test reactors around the world that currently use high-enriched uranium fuels. As part of this program, reactor experiments are being performed in the Advanced Test Reactor. It must be demonstrated that this fuel type exhibits mechanical integrity, geometric stability, and predictable behavior to high powers and high fission densities in order for it to be a viable fuel for qualification. This paper provides an overview of the microstructures observed at different regions of interest in fuel plates before and after irradiation for fuel samples that have been tested. These fuel plates were fabricated using laboratory-scale fabrication methods. Observations regarding how microstructural changes during irradiation may impact fuel performance are discussed.

SEM in situ MiniCantilever Beam Bending of U-10Mo/Zr/Al Fuel Elements

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

Mook, William; Baldwin, Jon K.; Martinez, Ricardo M.

2014-06-16

In this work, the fracture behavior of Al/Zr and Zr/dU-10Mo interfaces was measured via the minicantilever bend technique. The energy dissipation rates were found to be approximately 3.7-5 mj/mm 2 and 5.9 mj/mm 2 for each interface, respectively. It was found that in order to test the Zr/U-10Mo interface, location of the hinge of the cantilever was a key parameter. While this test could be adapted to hot cell use through careful alignment fixturing and measurement of crack lengths with an optical microscope (as opposed to SEM, which was used here out of convenience), machining of the cantilevers via MiniMillmore » in such a way as to locate the interfaces at the cantilever hinge, as well as proper placement of a femtosecond laser notch will continue to be key challenges in a hot cell environment.« less

Characterization of fission gas bubbles in irradiated U-10Mo fuel

DOE PAGES

Casella, Andrew M.; Burkes, Douglas E.; MacFarlan, Paul J.; ...

2017-06-06

A simple, repeatable method for characterization of fission gas bubbles in irradiated U-Mo fuels has been developed. This method involves mechanical potting and polishing of samples along with examination with a scanning electron microscope located outside of a hot cell. The commercially available software packages CellProfiler, MATLAB, and Mathematica are used to segment and analyze the captured images. The results are compared and contrasted. Finally, baseline methods for fission gas bubble characterization are suggested for consideration and further development.

Analysis and comparison of focused ion beam milling and vibratory polishing sample surface preparation methods for porosity study of U-Mo plate fuel for research and test reactors.

PubMed

Westman, Bjorn; Miller, Brandon; Jue, Jan-Fong; Aitkaliyeva, Assel; Keiser, Dennis; Madden, James; Tucker, Julie D

2018-07-01

Uranium-Molybdenum (U-Mo) low enriched uranium (LEU) fuels are a promising candidate for the replacement of high enriched uranium (HEU) fuels currently in use in a high power research and test reactors around the world. Contemporary U-Mo fuel sample preparation uses focused ion beam (FIB) methods for analysis of fission gas porosity. However, FIB possess several drawbacks, including reduced area of analysis, curtaining effects, and increased FIB operation time and cost. Vibratory polishing is a well understood method for preparing large sample surfaces with very high surface quality. In this research, fission gas porosity image analysis results are compared between samples prepared using vibratory polishing and FIB milling to assess the effectiveness of vibratory polishing for irradiated fuel sample preparation. Scanning electron microscopy (SEM) imaging was performed on sections of irradiated U-Mo fuel plates and the micrographs were analyzed using a fission gas pore identification and measurement script written in MatLab. Results showed that the vibratory polishing method is preferentially removing material around the edges of the pores, causing the pores to become larger and more rounded, leading to overestimation of the fission gas porosity size. Whereas, FIB preparation tends to underestimate due to poor micrograph quality and surface damage leading to inaccurate segmentations. Despite the aforementioned drawbacks, vibratory polishing remains a valid method for porosity analysis sample preparation, however, improvements should be made to reduce the preferential removal of material surrounding pores in order to minimize the error in the porosity measurements. Copyright © 2018 Elsevier Ltd. All rights reserved.

Four-point Bend Testing of Irradiated Monolithic U-10Mo Fuel

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

Rabin, B. H.; Lloyd, W. R.; Schulthess, J. L.

2015-03-01

This paper presents results of recently completed studies aimed at characterizing the mechanical properties of irradiated U-10Mo fuel in support of monolithic base fuel qualification. Mechanical properties were evaluated in four-point bending. Specimens were taken from fuel plates irradiated in the RERTR-12 and AFIP-6 Mk. II irradiation campaigns, and tests were conducted in the Hot Fuel Examination Facility (HFEF) at Idaho National Laboratory (INL). The monolithic fuel plates consist of a U-10Mo fuel meat covered with a Zr diffusion barrier layer fabricated by co-rolling, clad in 6061 Al using a hot isostatic press (HIP) bonding process. Specimens exhibited nominal (fresh)more » fuel meat thickness ranging from 0.25 mm to 0.64 mm, and fuel plate average burnup ranged from approximately 0.4 x 1021 fissions/cm 3 to 6.0 x 1021 fissions/cm 3. After sectioning the fuel plates, the 6061 Al cladding was removed by dissolution in concentrated NaOH. Pre- and post-dissolution dimensional inspections were conducted on test specimens to facilitate accurate analysis of bend test results. Four-point bend testing was conducted on the HFEF Remote Load Frame at a crosshead speed of 0.1 mm/min using custom-designed test fixtures and calibrated load cells. All specimens exhibited substantially linear elastic behavior and failed in a brittle manner. The influence of burnup on the observed slope of the stress-strain curve and the calculated fracture strength is discussed.« less

Identification of phases in the interaction layer between U-Mo-Zr/Al and U-Mo-Zr/Al-Si

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

Varela, C.L. Komar; Arico, S.F.; Mirandou, M.

Out-of-pile diffusion experiments were performed between U-7wt.% Mo-1wt.% Zr and Al or Al A356 (7,1wt.% Si) at 550 deg. C. In this work morphological characterization and phase identification on both interaction layer are presented. They were carried out by the use of different techniques: optical and scanning electron microscopy, X-Ray diffraction and WDS microanalysis. In the interaction layer U-7wt.% Mo-1wt.% Zr/Al, the phases UAl{sub 3}, UAl{sub 4}, Al{sub 20}Mo{sub 2}U and Al{sub 43}Mo{sub 4}U{sub 6} were identified. In the interaction layer U-7wt.% Mo-1wt.% Zr/Al A356, the phases U(Al, Si) with 25at.% Si and Si{sub 5}U{sub 3} were identified. This lastmore » phase, with a higher Si concentration, was identified with XRD Synchrotron radiation performed at the National Synchrotron Light Laboratory (LNLS), Campinas, Brasil. (author)« less

Microstructure of RERTR DU-Alloys Irradiated with Krypton Ions

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

J. Gan; D. Keiser; D. Wachs

2009-11-01

Fuel development for reduced enrichment research and test reactor (RERTR) program is tasked with the development of new low enrichment uranium fuels that can be employed to replace existing high enrichment uranium fuels currently used in many research and test reactors worldwide. Radiation stability of the interaction product formed at fuel-matrix interface has a strong impact on fuel performance. Three depleted uranium alloys are cast that consist of the following 5 phases of interest to be investigated: U(Si,Al)3, (U,Mo)(Si,Al)3, UMo2Al20, U6Mo4Al43 and UAl4. Irradiation of TEM disc samples with 500 keV Kr ions at 200?C to high doses up tomore » ~100 dpa were conducted using an intermediate voltage electron microscope equipped with an ion accelerator. The irradiated microstructure of the 5 phases is characterized using transmission electron microscopy. The results will be presented and the implication of the observed irradiated microstructure on the fuel performance will be discussed.« less

Molybdenum-99 production calculation analysis of SAMOP reactor based on thorium nitrate fuel

NASA Astrophysics Data System (ADS)

Syarip; Togatorop, E.; Yassar

2018-03-01

SAMOP (Subcritical Assembly for Molybdenum-99 Production) has the potential to use thorium as fuel to produce 99Mo after modifying the design, but the production performance has not been discovered yet. A study needs to be done to obtain the correlation between 99Mo production with the mixed fuel composition of uranium and with SAMOP power on the modified SAMOP design. The study aims to obtain the production of 99Mo based thorium nitrate fuel on SAMOP’s modified designs. Monte Carlo N-Particle eXtended (MCNPX) is required to simulate the operation of the assembly by varying the composition of the uranium-thorium nitrate mixed fuel, geometry and power fraction on the SAMOP modified designs. The burnup command on the MCNPX is used to confirm the 99Mo production result. The assembly is simulated to operate for 6 days with subcritical neutron multiplication factor (keff = 0.97-0.99). The neutron multiplication factor of the modified design (keff) is 0.97, the activity obtained from 99Mo is 18.58 Ci at 1 kW power operation.

Report on the Synchrotron Characterization of U-Mo and U-Zr Alloys and the Modeling Results

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

Okuniewski, Maria A.; Ganapathy, Varsha; Hamilton, Brenden

2016-09-01

ABSTRACT Uranium-molybdenum (U-Mo) and uranium-zirconium (U-Zr) are two promising fuel candidates for nuclear transmutation reactors which burn long-lived minor actinides and fission products within fast spectrum reactors. The objectives of this research are centered on understanding the early stages of fuel performance through the examination of the irradiation induced microstructural changes in U-Zr and U-Mo alloys subjected to low neutron fluences. Specimens that were analyzed include those that were previously irradiated in the Advanced Test Reactor at INL. This most recent work has focused on a sub-set of the irradiated specimens, specifically U-Zr and U-Mo alloys that were irradiated tomore » 0.01 dpa at temperatures ranging from (150-800oC). These specimens were analyzed with two types of synchrotron techniques, including X-ray absorption fine structure and X-ray diffraction. These techniques provide non-destructive microstructural analysis, including phase identification and quantitation, lattice parameters, crystallite sizes, as well as bonding, structure, and chemistry. Preliminary research has shown changes in the phase fractions, crystallite sizes, and lattice parameters as a function of irradiation and temperature. Future data analyses will continue to explore these microstructural changes.« less

Copper nanoparticle interspersed MoS2 nanoflowers with enhanced efficiency for CO2 electrochemical reduction to fuel.

PubMed

Shi, Guodong; Yu, Luo; Ba, Xin; Zhang, Xiaoshu; Zhou, Jianqing; Yu, Ying

2017-08-15

Electrocatalytic conversion of carbon dioxide (CO 2 ) has been considered as an ideal method to simultaneously solve the energy crisis and environmental issue around the world. In this work, ultrasmall Cu nanoparticle interspersed flower-like MoS 2 was successfully fabricated via a facile microwave hydrothermal method. The designed optimal hierarchical Cu/MoS 2 composite not only exhibited remarkably enhanced electronic conductivity and specific surface area but also possessed improved CO 2 adsorption capacity, resulting in a significant increase in overall faradaic efficiency and a 7-fold augmentation of the faradaic efficiency of CH 4 in comparison with bare MoS 2 . In addition, the Cu/MoS 2 composite had superior stability with high efficiency retained for 48 h in the electrochemical process. It is anticipated that the designed Cu/MoS 2 composite electrocatalyst may provide new insights for transition metal sulfides and non-noble particles applied to CO 2 reduction.

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

Swelling of U-7Mo/Al-Si dispersion fuel plates under irradiation - Non-destructive analysis of the AFIP-1 fuel plates

NASA Astrophysics Data System (ADS)

Wachs, D. M.; Robinson, A. B.; Rice, F. J.; Kraft, N. C.; Taylor, S. C.; Lillo, M.; Woolstenhulme, N.; Roth, G. A.

2016-08-01

Extensive fuel-matrix interactions leading to plate pillowing have proven to be a significant impediment to the development of a suitable high density low-enriched uranium molybdenum alloy (U-Mo) based dispersion fuel for high power applications in research reactors. The addition of silicon to the aluminum matrix was previously demonstrated to reduce interaction layer growth in mini-plate experiments. The AFIP-1 project involved the irradiation, in-canal examination, and post-irradiation examination of two fuel plates. The irradiation of two distinct full size, flat fuel plates (one using an Al-2wt%Si matrix and the other an Al-4043 (∼4.8 wt% Si) matrix) was performed in the INL ATR reactor in 2008-2009. The irradiation conditions were: ∼250 W/cm2 peak Beginning Of Life (BOL) power, with a ∼3.5e21 f/cm3 peak burnup. The plates were successfully irradiated and did not show any pillowing at the end of the irradiation. This paper reports the results and interpretation of the in-canal and post-irradiation non-destructive examinations that were performed on these fuel plates. It further compares additional PIE results obtained on fuel plates irradiated in contemporary campaigns in order to allow a complete comparison with all results obtained under similar conditions. Except for a brief indication of accelerated swelling early in the irradiation of the Al-2Si plate, the fuel swelling is shown to evolve linearly with the fission density through the maximum burnup.

The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

NASA Astrophysics Data System (ADS)

Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

2017-04-01

The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world's highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding during fabrication and are enhanced during irradiation. One aspect of fuel development and qualification is to demonstrate an appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding and Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 °C). The mechanisms responsible for fission gas release events are discussed.

The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

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

Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

2017-04-01

The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world’s highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form during fabrication and are enhanced during irradiation between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding. One aspect of fuel development and qualification is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding andmore » Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 oC). The mechanisms responsible for fission gas release events are discussed.« less

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

Hu, Shenyang; Setyawan, Wahyu; Joshi, Vineet V.

Xe gas bubble superlattice formation is observed in irradiated uranium–10 wt% molybdenum (U10Mo) fuels. However, the thermodynamic properties of the bubbles (the relationship among bubble size, equilibrium Xe concentration, and bubble pressure) and the mechanisms of bubble growth and superlattice formation are not well known. In this work, molecular dynamics is used to study these properties and mechanisms. The results provide important inputs for quantitative mesoscale models of gas bubble evolution and fuel performance. In the molecular dynamics simulations, the embedded-atom method (EAM) potential of U10Mo-Xe (Smirnova et al. 2013) is employed. Initial gas bubbles with low Xe concentration aremore » generated in a U10Mo single crystal. Then Xe atom atoms are continuously added into the bubbles, and the evolution of pressure and dislocation emission around the bubbles is analyzed. The relationship between pressure, equilibrium Xe concentration, and radius of the bubbles is established. It was found that the gas bubble growth is accompanied by partial dislocation emission, which results in a star-shaped dislocation structure and an anisotropic stress field. The emitted partial dislocations have a Burgers vector along the <111> direction and a slip plane of (11-2). Dislocation loop punch-out was not observed. A tensile stress was found along <110> directions around the bubble, favoring the nucleation and formation of a face-centered cubic bubble superlattice in body-centered cubic U10Mo fuels.« less

Image fusion of Secondary Ion Mass Spectrometry and Energy-dispersive X-Ray Spectroscopy data for the characterization of uranium-molybdenum fuel foils

NASA Astrophysics Data System (ADS)

Willingham, David; Naes, Benjamin E.; Tarolli, Jay G.; Schemer-Kohrn, Alan; Rhodes, Mark; Dahl, Michael; Guzman, Anthony; Burkes, Douglas E.

2018-01-01

Uranium-molybdenum (U-Mo) monolithic fuels represent one option for converting civilian research and test reactors operating with high enriched uranium (HEU) to low enriched uranium (LEU), effectively reducing the threat of nuclear proliferation world-wide. However, processes associated with fabrication of U-Mo monolithic fuels result in regions of elemental heterogeneity, observed as bands traversing the cross-section of representative samples. Isotopic variations (e.g., 235U and 238U) could also be introduced because of associated processing steps, particularly since HEU feedstock is melted with natural or depleted uranium diluent to produce LEU. This study demonstrates the utility of correlative analysis of Energy-Dispersive X-ray Spectroscopy (EDS) and Secondary Ion Mass Spectrometry (SIMS) with their image data streams using image fusion, resulting in a comprehensive microanalytical characterization toolbox. Elemental and isotopic measurements were made on a sample from the Advanced Test Reactor (ATR) Full-sized plate In-center flux trap Position (AFIP)-7 experiment and compared to previous optical and electron microscopy results. The image fusion results are characteristic of SIMS isotopic maps, but with the spatial resolution of EDS images and, therefore, can be used to increase the effective spatial resolution of the SIMS imaging results to better understand homogeneity or heterogeneity that persists because of processing selections. Visual inspection using the image fusion methodology indicated slight variations in the 235U/238U ratio and quantitative analysis using the image intensities across several FoVs revealed an average 235U atom percent value of 17.9 ± 2.4%, which was indicative of a non-uniform U isotopic distribution in the area sampled. Further development of this capability is useful for understanding the connections between the properties of LEU fuel alternatives and the ability to predict performance under irradiation.

Estimation of weekly 99Mo production by AHR 200 kW

NASA Astrophysics Data System (ADS)

Siregar, I. H.; Suharyana; Khakim, A.; Siregar, D.; Frida, A. R.

2016-11-01

The estimation of weekly 99Mo production by AHR 200 kW fueled with Low Enriched Uranium Uranyl Nitrate solution has been simulated by using MCNPX computer code. We have employed the AHR design of Babcock & Wilcox Medical Isotope Production System with 9Be Reflector and Stainless steel vessel. We found that when the concentration of uranium in the fresh fuel was 108 gr U/L of UO2(NO3)2 fuel solution, the multiplication factor was 1.0517. The 99Mo concentration reached saturated at tenth day operation. The AHR can produce approximately 1.96×103 6-day-Ci weekly.

Kr ion irradiation study of the depleted-uranium alloys

NASA Astrophysics Data System (ADS)

Gan, J.; Keiser, D. D.; Miller, B. D.; Kirk, M. A.; Rest, J.; Allen, T. R.; Wachs, D. M.

2010-12-01

Fuel development for the reduced enrichment research and test reactor (RERTR) program is tasked with the development of new low enrichment uranium nuclear fuels that can be employed to replace existing high enrichment uranium fuels currently used in some research reactors throughout the world. For dispersion type fuels, radiation stability of the fuel-cladding interaction product has a strong impact on fuel performance. Three depleted-uranium alloys are cast for the radiation stability studies of the fuel-cladding interaction product using Kr ion irradiation to investigate radiation damage from fission products. SEM analysis indicates the presence of the phases of interest: U(Al, Si) 3, (U, Mo)(Al, Si) 3, UMo 2Al 20, U 6Mo 4Al 43 and UAl 4. Irradiations of TEM disc samples were conducted with 500 keV Kr ions at 200 °C to ion doses up to 2.5 × 10 19 ions/m 2 (˜10 dpa) with an Kr ion flux of 10 16 ions/m 2/s (˜4.0 × 10 -3 dpa/s). Microstructural evolution of the phases relevant to fuel-cladding interaction products was investigated using transmission electron microscopy.

Paleo-environmental conditions of the Early Cambrian Niutitang Formation in the Fenggang area, the southwestern margin of the Yangtze Platform, southern China: Evidence from major elements, trace elements and other proxies

NASA Astrophysics Data System (ADS)

Li, Jin; Tang, Shuheng; Zhang, Songhang; Xi, Zhaodong; Yang, Ning; Yang, Guoqiao; Li, Lei; Li, Yanpeng

2018-06-01

The Precambrian/Cambrian transition was a key time in Earth history, especially for marine biological evolution and oceanic chemistry. The redox-stratification with oxic shallow water and anoxic (even euxinic) deeper water in the Early Cambrian Yangtze Sea, which gradually became completely oxygenated, has been suggested as a possible trigger for the "Cambrian explosion" of biological diversity. However, for some areas in northern Guizhou where the exploration and research are lacking, identifying this pattern of redox-stratification by paleo-environmental analysis from borehole data is still in need. Here, we report a remarkable variation range in trace elements (Mo, V, U, Ni, Th, Co, Sc, Zn and Cu), molar Corg:P ratios and pyrite morphology from 27 core samples from one new drill hole (XY1, located in the Fenggang area, northern Guizhou) on the Yangtze Platform, South China. High levels of Ba (from 3242 ppm to 33,800 ppm) and total organic carbon (TOC; from 4% to 9.36%) in 15 core samples in the Lower Member (LM) of the Niutitang Formation indicated elevated primary productivity in the study area. Redox change was recorded based on enrichment factors (EFs) for RSTEs (Mo, U, and V), redox proxies (V/(V + Ni), Ni/Co, V/Sc and Th/U), Corg:P ratios and particle size of framboidal pyrite. These signatures demonstrate that the LM was deposited under anoxic conditions with sulfidic episodes, whereas the Upper Member (UM) of the Niutitang Formation was deposited under suboxic/oxic conditions with intermittently anoxic episodes. Mo/TOC ratios (from 3.72 to 39.86, mean 18.76) suggest weak-moderate water mass restriction. Mo-U covariation patterns (strong but variable enrichment of Mo and U; MoEF ranging from 31.45 to 257.97; UEF ranging from 4.68 to 39.07) in the LM show alternation of particulate shuttling and redox conditions occurred in the Early Cambrian Yangtze Sea, whereas Mo-U covariation patterns (moderate Mo enrichment but depletion or non-enrichment of U; mean MoEF: 7.29; mean UEF: 0.95) in the UM may indicate the combined influence of particulate shuttling and diagenetic diffusion of U via bioactivities, which result in low U values and an anoxic signature from frambiodal pyrite particle size (mean: 4.556 μm; median: 4.41 μm). Additionally, excess Ba (Baxs) concentration (33,800 ppm and 32,500 ppm) and association patterns of trace-metal enrichment in the LM indicate the existence of submarine hydrothermal events. In addition, during deposition of the UM, bioactivities indicated by Mo-U systematics and oxic conditions indicated by redox sensitive trace elements (RSTEs) and multiple-proxies, may be a cause of biological diversification recorded in the Early Cambrian. Finally, data in this record a progressive transition from anoxic bottom waters with euxinic episodes to overwhelming oxic conditions during Early Cambrian.

Microstructure of the irradiated U 3Si 2/Al silicide dispersion fuel

NASA Astrophysics Data System (ADS)

Gan, J.; Keiser, D. D.; Miller, B. D.; Jue, J.-F.; Robinson, A. B.; Madden, J. W.; Medvedev, P. G.; Wachs, D. M.

2011-12-01

The silicide dispersion fuel of U 3Si 2/Al is recognized as the best performance fuel for many nuclear research and test reactors with up to 4.8 gU/cm 3 fuel loading. An irradiated U 3Si 2/Al dispersion fuel ( 235U ˜ 75%) from the high-flux side of a fuel plate (U0R040) from the Reduced Enrichment for Research and Test Reactors (RERTR)-8 test was characterized using transmission electron microscopy (TEM). The fuel was irradiated in the Advanced Test Reactor (ATR) for 105 days. The average irradiation temperature and fission density of the U 3Si 2 fuel particles for the TEM sample are estimated to be approximately 110 °C and 5.4 × 10 27 f/m 3. The characterization was performed using a 200-kV TEM. The U/Si ratio for the fuel particle and (Si + Al)/U for the fuel-matrix-interaction layer are approximately 1.1 and 4-10, respectively. The estimated average diameter, number density and volume fraction for small bubbles (<1 μm) in the fuel particle are ˜94 nm, 1.05 × 10 20 m -3 and ˜11%, respectively. The results and their implication on the performance of the U 3Si 2/Al silicide dispersion fuel are discussed.

An aerosol particle containing enriched uranium encountered in the remote upper troposphere.

PubMed

Murphy, D M; Froyd, K D; Apel, E; Blake, D; Blake, N; Evangeliou, N; Hornbrook, R S; Peischl, J; Ray, E; Ryerson, T B; Thompson, C; Stohl, A

2018-04-01

We describe a submicron aerosol particle sampled at an altitude of 7 km near the Aleutian Islands that contained a small percentage of enriched uranium oxide. 235 U was 3.1 ± 0.5% of 238 U. During twenty years of aircraft sampling of millions of particles in the global atmosphere, we have rarely encountered a particle with a similarly high content of 238 U and never a particle with enriched 235 U. The bulk of the particle consisted of material consistent with combustion of heavy fuel oil. Analysis of wind trajectories and particle dispersion model results show that the particle could have originated from a variety of areas across Asia. The source of such a particle is unclear, and the particle is described here in case it indicates a novel source where enriched uranium was dispersed. Published by Elsevier Ltd.

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

Hill, Mary Ann; Dombrowski, David E.; Clarke, Kester Diederik

U-10 wt. % Mo (U-10Mo) alloys are being developed as low enrichment monolithic fuel for the CONVERT program. Optimization of processing for the monolithic fuel is being pursued with the use of electrical discharge machining (EDM) under CONVERT HPRR WBS 1.2.4.5 Optimization of Coupon Preparation. The process is applicable to manufacturing experimental fuel plate specimens for the Mini-Plate-1 (MP-1) irradiation campaign. The benefits of EDM are reduced machining costs, ability to achieve higher tolerances, stress-free, burr-free surfaces eliminating the need for milling, and the ability to machine complex shapes. Kerf losses are much smaller with EDM (tenths of mm) comparedmore » to conventional machining (mm). Reliable repeatability is achievable with EDM due to its computer-generated machining programs.« less

Technical assumption for Mo-99 production in the MARIA reactor. Feasibility study

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

Jaroszewicz, J.; Pytel, K.; Dabkowski, L.

2008-07-15

The main objective of U-235 irradiation is to obtain the Tc-99m isotope which is widely used in the domain of medical diagnostics. The decisive factor determining its availability, despite its short life time, is a reaction of radioactive decay of Mo-99 into Tc- 99m. One of the possible sources of molybdenum can be achieved in course of the U-235 fission reaction. The paper presents activities and the calculations results obtained upon the feasibility study on irradiation of U-235 targets for production of molybdenum in the MARIA reactor. The activities including technical assumption were focused on performing calculation for modelling ofmore » the target and irradiation device as well as adequate equipment and tools for processing in reactor. It has been assumed that the basic component of fuel charge is an aluminium cladded plate with dimensions of 40x230x1.45 containing 4.7 g U-235. The presumed mode of the heat removal generated in the fuel charge of the reactor primary cooling circuit influences the construction of installation to be used for irradiation and the technological instrumentation. The outer channel construction for irradiation has to be identical as the standard fuel channel construction of the MARIA reactor. It enables to use the existing slab and reactor mounting sockets for the fastening of the molybdenum channel as well as the cooling water delivery system. The measurement of water temperature cooling a fuel charge and control of water flow rate in the channel can also be carried out be means of the standard instrumentation of the reactor. (author)« less

Processing of U-2.5Zr-7.5Nb and U-3Zr-9Nb alloys by sintering process

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

Dos Santos, A. M. M.; Ferraz, W. B.; Lameiras, F. S.

2012-07-01

To minimize the risk of nuclear proliferation, there is worldwide interest in reducing fuel enrichment of research and test reactors. To achieve this objective while still guaranteeing criticality and cycle length requirements, there is need of developing high density uranium metallic fuels. Alloying elements such as Zr, Nb and Mo are added to uranium to improve fuel performance in reactors. In this context, the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) is developing the U-2.5Zr-7.5Nb and U-3Zr-9Nb (weight %) alloys by the innovative process of sintering that utilizes raw materials in the form of powders. The powders were pressed atmore » 400 MPa and then sintered under a vacuum of about 1x10{sup -4} Torr at temperatures ranging from 1050 deg. to 1500 deg.C. The densities of the alloys were measured geometrically and by hydrostatic method and the phases identified by X ray diffraction (XRD). The microstructures of the pellets were observed by scanning electron microscopy (SEM) and the alloying elements were analyzed by energy dispersive X-ray spectroscopy (EDS). The results obtained showed the fuel density to slightly increase with the sintering temperature. The highest density achieved was approximately 80% of theoretical density. It was observed in the pellets a superficial oxide layer formed during the sintering process. (authors)« less

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

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

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

2015-03-01

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

ATR LEU fuel and burnable absorber neutronics performance optimization by fuel meat thickness variation

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

Chang, G.S.

2008-07-15

The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuelmore » cycle burnup comparison analysis. Using the current HEU U-235 enrichment of 93.0 % as a baseline, an analysis can be performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff between the HEU core th and the LEU core versus effective full power days (EFPD). The MCNP ATR 1/8th core model will be used to optimize the U-235 loading in the LEU core, such that the differences in K-eff and heat flux profile between the HEU and LEU core can be minimized. The depletion methodology MCWO was used to calculate K-eff versus EFPDs in this paper. The MCWO-calculated results for the LEU cases with foil (U-10Mo) types demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar to the reference ATR HEU case. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm. In this work, the proposed LEU (U-10Mo) core conversion case with a nominal fuel meat thickness of 0.381 mm and the same U-235 enrichment (19.7 wt%) can be used to optimize the radial heat flux profile by varying the fuel meat thickness from 0.191 mm (7.5 mil) to 0.343 mm (13.5 mil) at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19). In addition, 0.8g of a burnable absorber, Boron-10, was added in the inner and outer plates to reduce the initial excess reactivity, and the inner/outer heat flux more effectively. The optimized LEU relative radial fission heat flux profile is bounded by the reference ATR HEU case. However, to demonstrate that the LEU core fuel cycle performance can meet the Updated Final Safety Analysis Report (UFSAR) safety requirements, additional studies will be necessary to evaluate and compare safety parameters such as void reactivity and Doppler coefficients, control components worth (outer shim control cylinders, safety rods and regulating rod), and shutdown margins between the HEU and LEU cores. (author)« less

NEUTRONIC REACTOR FUEL ELEMENT

DOEpatents

Shackleford, M.H.

1958-12-16

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

Concept Feasibility Report for Electroplating Zirconium onto Uranium Foil - Year 2

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

Coffey, Greg W.; Meinhardt, Kerry D.; Joshi, Vineet V.

2015-03-01

The Fuel Fabrication Capability within the U.S. High Performance Research Reactor Conversion Program is funded through the National Nuclear Security Administration (NNSA) NA-26 (Office of Material Management and Minimization). An investigation was commissioned to determine the feasibility of using electroplating techniques to apply a coating of zirconium onto depleted uranium/molybdenum alloy (U-10Mo). Electroplating would provide an alternative method to the existing process of hot roll-bonding zirconium foil onto the U-10Mo fuel foil during the fabrication of fuel elements for high-performance research reactors. The objective of this research was to develop a reproducible and scalable plating process that will produce amore » uniform, 25 μm thick zirconium metal coating on U-10Mo foil. In previous work, Pacific Northwest National Laboratory (PNNL) established a molten salt electroplating apparatus and protocol to plate zirconium metal onto molybdenum foil (Coffey 2015). During this second year of the research, PNNL furthered this work by moving to the U-10Mo alloy system (90 percent uranium:10 percent molybdenum). The original plating apparatus was disassembled and re-assembled in a laboratory capable of handling low-level radioactive materials. Initially, the work followed the previous year’s approach, and the salt bath composition was targeted at the eutectic composition (LiF:NaF:ZrF4 = 26:37:37 mol%). Early results indicated that the formation of uranium fluoride compounds would be problematic. Other salt bath compositions were investigated in order to eliminate the uranium fluoride production (LiF:NaF = 61:39 mol% and LiF:NaF:KF = 46.5:11.5:42 mol% ). Zirconium metal was used as the crucible for the molten salt. Three plating methods were used—isopotential, galvano static, and pulsed plating. The molten salt method for zirconium metal application provided high-quality plating on molybdenum in PNNL’s previous work. A key advantage of this approach is that plating can be performed under conditions that would greatly reduce the quantity of intermetallics that form at the interface between the zirconium and U-10Mo; unlike roll bonding, the molten salt plating approach would allow for complete coverage of the U-10Mo foil with zirconium. When utilizing the experimental parameters developed for zirconium plating onto molybdenum, a uranium fluoride reaction product was formed at the Zr/U-10Mo interface. By controlling the initial plating potential, the uranium fluoride could be prevented; however, the targeted zirconium thickness (25 ±12.5 μm) could not be achieved while maintaining 100% coverage.« less

Interdiffusion behaviors in doped molybdenum uranium and aluminum or aluminum silicon dispersion fuels: Effects of the microstructure

NASA Astrophysics Data System (ADS)

Allenou, J.; Tougait, O.; Pasturel, M.; Iltis, X.; Charollais, F.; Anselmet, M. C.; Lemoine, P.

2011-09-01

Si addition to Al is considered as a promising route to reduce (U,Mo)-Al interaction kinetics, due to its accumulation in the interaction layer, yielding the formation of silicide phases. The (U,Mo) alloy microstructure, and especially its homogenization state, could play a role on this accumulation process. The addition of a third element in γ(U,Mo) could also influence diffusion mechanisms of Al and Si. These two parameters were studied by means of diffusion couple experiments by joining γU based alloys with Al and (Al,Si) alloy. Chemical elements X added into γ(U,Mo) were thoroughly chosen on the following criteria: (i) the potential solubility of the alloying element into the γ(U,Mo) matrix, (ii) its capability to form the ternary aluminides based on the CeCr 2Al 20 and Ho 6Mo 4Al 43 - types, and (iii) the feasibility to control the microstructure of the alloys. On this basis, a test matrix is defined. It concerns γ(U80,Mo15,X5) alloys (in at.%) with X = Y, Cu, Zr, Ti or Cr. These alloys were homogenized and coupled with Al or (Al,Si) alloy. Results evidenced, first, the importance of the state of homogenization of the γ(U,Mo) binary alloy on interaction processes with (Al,Si) alloy, and the benefit on the diffusion of Si through the interaction layer, as observed on the elementary concentration profiles, when the third element X has some solubility into γ(U,Mo) alloy.

Grain boundary engineering to control the discontinuous precipitation in multicomponent U10Mo alloy

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

Devaraj, Arun; Kovarik, Libor; Kautz, Elizabeth

Grain boundaries in metallic alloys often play a crucial role, not only in determining the mechanical properties or thermal stability of alloys, but also in dictating the phase transformation kinetics during thermomechanical processing. We demonstrate that locally stabilized structure and compositional segregation at grain boundaries—“grain boundary complexions”—in a complex multicomponent alloy can be modified to influence the kinetics of cellular transformation during subsequent thermomechanical processing. Using aberration-corrected scanning transmission electron microscopy and atom probe tomography analysis of a metallic nuclear fuel highly relevant to worldwide nuclear non-proliferation efforts —uranium-10 wt% molybdenum (U-10Mo) alloy, new evidence for the existence of grainmore » boundary complexion is provided. We then modified the concentration of impurities dissolved in Υ-UMo grain interiors and/or segregated to Υ-UMo grain boundaries by changing the homogenization treatment, and these effects were used used to retard the kinetics of cellular transformation during subsequent sub-eutectoid annealing in this U-10-Mo alloy during sub-eutectoid annealing. Thus, this work provided insights on tailoring the final microstructure of the U-10Mo alloy, which can potentially improve the irradiation performance of this important class of alloy fuels.« less

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

A model for recovery of scrap monolithic uranium molybdenum fuel by electrorefining

NASA Astrophysics Data System (ADS)

Van Kleeck, Melissa A.

The goal of the Reduced Enrichment for Research and Test Reactors program (RERTR) is toreduce enrichment at research and test reactors, thereby decreasing proliferation risk at these facilities. A new fuel to accomplish this goal is being manufactured experimentally at the Y12 National Security Complex. This new fuel will require its own waste management procedure,namely for the recovery of scrap from its manufacture. The new fuel is a monolithic uraniummolybdenum alloy clad in zirconium. Feasibility tests were conducted in the Planar Electrode Electrorefiner using scrap U-8Mo fuel alloy. These tests proved that a uranium product could be recovered free of molybdenum from this scrap fuel by electrorefining. Tests were also conducted using U-10Mo Zr clad fuel, which confirmed that product could be recovered from a clad version of this scrap fuel at an engineering scale, though analytical results are pending for the behavior of Zr in the electrorefiner. A model was constructed for the simulation of electrorefining the scrap material produced in the manufacture of this fuel. The model was implemented on two platforms, Microsoft Excel and MatLab. Correlations, used in the model, were developed experimentally, describing area specific resistance behavior at each electrode. Experiments validating the model were conducted using scrap of U-10Mo Zr clad fuel in the Planar Electrode Electrorefiner. The results of model simulations on both platforms were compared to experimental results for the same fuel, salt and electrorefiner compositions and dimensions for two trials. In general, the model demonstrated behavior similar to experimental data but additional refinements are needed to improve its accuracy. These refinements consist of a function for surface area at anode and cathode based on charge passed. Several approximations were made in the model concerning areas of electrodes which should be replaced by a more accurate function describing these areas.

Experimental physics characteristics of a heavy-metal-reflected fast-spectrum critical assembly

NASA Technical Reports Server (NTRS)

Heneveld, W. H.; Paschall, R. K.; Springer, T. H.; Swanson, V. A.; Thiele, A. W.; Tuttle, R. J.

1971-01-01

A zero-power critical assembly was designed, constructed, and operated for the purpose of conducting a series of benchmark experiments dealing with the physics characteristics of a UN-fueled, Li-7 cooled, Mo-reflected, drum-controlled compact fast reactor for use with a space-power electric conversion system. The experimental program consisted basically of measuring the differential neutron spectra and the changes in critical mass that accompanied the stepwise addition of (Li-7)3N, Hf, Ta, and W to a basic core fueled with U metal in a pin-type Ta honeycomb structure. In addition, experimental results were obtained on power distributions, control characteristics, neutron lifetime, and reactivity worths of numerous absorber, structural, and scattering materials.

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

Liang, Linyun; Mei, Zhi -Gang; Kim, Yeon Soo

A mesoscale model is developed by integrating the rate theory and phase-field models and is used to study the fission-induced recrystallization in U-7Mo alloy. The rate theory model is used to predict the dislocation density and the recrystallization nuclei density due to irradiation. The predicted fission rate and temperature dependences of the dislocation density are in good agreement with experimental measurements. This information is used as input for the multiphase phase-field model to investigate the fission-induced recrystallization kinetics. The simulated recrystallization volume fraction and bubble induced swelling agree well with experimental data. The effects of the fission rate, initial grainmore » size, and grain morphology on the recrystallization kinetics are discussed based on an analysis of recrystallization growth rate using the modified Avrami equation. Here, we conclude that the initial microstructure of the U-Mo fuels, especially the grain size, can be used to effectively control the rate of fission-induced recrystallization and therefore swelling.« less

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

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

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

1997-08-01

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

Neutron and hard X-ray diffraction studies of the isothermal transformation kinetics in the research reactor fuel candidate U–8 wt%Mo

PubMed Central

Säubert, Steffen; Jungwirth, Rainer; Zweifel, Tobias; Hofmann, Michael; Hoelzel, Markus; Petry, Winfried

2016-01-01

Exposing uranium–molybdenum alloys (UMo) retained in the γ phase to elevated temperatures leads to transformation reactions during which the γ-UMo phase decomposes into the thermal equilibrium phases, i.e. U2Mo and α-U. Since α-U is not suitable for a nuclear fuel exposed to high burn-up, it is necessary to retain the γ-UMo phase during the production process of the fuel elements for modern high-performance research reactors. The present work deals with the isothermal transformation kinetics in U–8 wt%Mo alloys for temperatures between 673 and 798 K and annealing durations of up to 48 h. Annealed samples were examined at room temperature using either X-ray or neutron diffraction to determine the phase composition after thermal treatment, and in situ annealing studies disclosed the onset of phase decomposition. While for temperatures of 698 and 673 K the start of decomposition is delayed, for higher temperatures the first signs of transformation are already observable within 3 h of annealing. The typical C-shaped curves in a time–temperature–transformation (TTT) diagram for both the start and the end of phase decomposition could be determined in the observed temperature regime. Therefore, a revised TTT diagram for U–8 wt%Mo between 673 and 798 K and annealing durations of up to 48 h is proposed. PMID:27275139

Processing and microstructural characterisation of a UO2-based ceramic for disposal studies on spent AGR fuel

NASA Astrophysics Data System (ADS)

Hiezl, Z.; Hambley, D. I.; Padovani, C.; Lee, W. E.

2015-01-01

Preparation and characterisation of a Simulated Spent Nuclear Fuel (SIMFuel), which replicates the chemical state and microstructure of Spent Nuclear Fuel (SNF) discharged from a UK Advanced Gas-cooled Reactor (AGR) after a cooling time of 100 years is described. Given the relatively small differences in radionuclide inventory expected over longer time periods, the SIMFuel studied in this work is expected to be also representative of spent fuel after significantly longer periods (e.g. 1000 years). Thirteen stable elements were added to depleted UO2 and sintered to simulate the composition of fuel pellets after burn-ups of 25 and 43 GWd/tU and, as a reference, pure UO2 pellets were also investigated. The fission product distribution was calculated using the FISPIN code provided by the UK National Nuclear Laboratory. SIMFuel pellets were up to 92% dense and during the sintering process in H2 atmosphere Mo-Ru-Rh-Pd metallic precipitates and grey-phase ((Ba, Sr)(Zr, RE) O3 oxide precipitates) formed within the UO2 matrix. These secondary phases are present in real PWR and AGR SNF. Metallic precipitates are generally spherical and have submicron particle size (0.8 ± 0.7 μm). Spherical oxide precipitates in SIMFuel measured up to 30 μm in diameter, but no data were available in the public domain to compare this to AGR SNF. The grain size of actual AGR SNF (∼ 3-30 μm) is larger than that measured in AGR SIMFuel (∼ 2-5 μm).

ATR LEU Fuel and Burnable Absorber Neutronics Performance Optimization by Fuel Meat Thickness Variation

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

G. S. Chang

2007-09-01

The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuelmore » cycle burnup comparison analysis. Using the current HEU U 235 enrichment of 93.0 % as a baseline, an analysis can be performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff between the HEU core and the LEU core versus effective full power days (EFPD). The MCNP ATR 1/8th core model will be used to optimize the U-235 loading in the LEU core, such that the differences in K-eff and heat flux profile between the HEU and LEU core can be minimized. The depletion methodology MCWO was used to calculate K-eff versus EFPDs in this paper. The MCWO-calculated results for the LEU cases with foil (U-10Mo) types demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar to the reference ATR HEU case. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm. In this work, the proposed LEU (U-10Mo) core conversion case with a nominal fuel meat thickness of 0.508 mm and the same U-235 enrichment (15.5 wt%) can be used to optimize the radial heat flux profile by varying the fuel plate thickness from 0.254 to 0.457 mm at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19). In addition, a 0.7g of burnable absorber Boron-10 was added in the inner and outer plates to reduce the initial excess reactivity, and the inner/outer heat flux more effectively. The optimized LEU relative radial fission heat flux profile is bounded by the reference ATR HEU case. However, to demonstrate that the LEU core fuel cycle performance can meet the Updated Final Safety Analysis Report (UFSAR) safety requirements, additional studies will be necessary to evaluate and compare safety parameters such as void reactivity and Doppler coefficients, control components worth (outer shim control cylinders, safety rods and regulating rod), and shutdown margins between the HEU and LEU cores.« less

Fuel Thermo-physical Characterization Project. Fiscal Year 2014 Final Report

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

Burkes, Douglas; Casella, Andrew M.; Buck, Edgar C.

2015-03-15

The Office of Material Management and Minimization (M3) Reactor Conversion Fuel Thermo-Physical Characterization Project at Pacific Northwest National Laboratory (PNNL) was tasked with using PNNL facilities and processes to receive irradiated low enriched uranium–molybdenum (LEU-Mo) fuel plate samples and perform analysis in support of the M3 Reactor Conversion Program. This work is in support of the M3 Reactor Conversion Fuel Development Pillar that is managed by Idaho National Laboratory. The primary research scope was to determine the thermo-physical properties as a function of temperature and burnup. Work conducted in Fiscal Year (FY) 2014 complemented measurements performed in FY 2013 onmore » four additional irradiated LEU-Mo fuel plate samples. Specifically, the work in FY 2014 investigated the influence of different processing methods on thermal property behavior, the absence of aluminum alloy cladding on thermal property behavior for additional model validation, and the influence of higher operating surface heat flux / more aggressive irradiation conditions on thermal property behavior. The model developed in FY 2013 and refined in FY 2014 to extract thermal properties of the U-Mo alloy from the measurements conducted on an integral fuel plate sample (i.e., U-Mo alloy with a thin Zr coating and clad in AA6061) continues to perform very well. Measurements conducted in FY 2014 on samples irradiated under similar conditions compare well to measurements performed in FY 2013. In general, there is no gross influence of fabrication method on thermal property behavior, although the difference in LEU-Mo foil microstructure does have a noticeable influence on recrystallization of grains during irradiation. Samples irradiated under more aggressive irradiation conditions, e.g., higher surface heat flux, revealed lower thermal conductivity when compared to samples irradiated at moderate surface heat fluxes, with the exception of one sample. This report documents thermal property measurements conducted in FY 2014 and compares results to values obtained from literature and measurements performed in FY 2013, where applicable, along with appropriate discussion.« less

Interdiffusion, Intrinsic Diffusion, Atomic Mobility, and Vacancy Wind Effect in γ(bcc) Uranium-Molybdenum Alloy

NASA Astrophysics Data System (ADS)

Huang, Ke; Keiser, Dennis D.; Sohn, Yongho

2013-02-01

U-Mo alloys are being developed as low enrichment uranium fuels under the Reduced Enrichment for Research and Test Reactor (RERTR) Program. In order to understand the fundamental diffusion behavior of this system, solid-to-solid pure U vs Mo diffusion couples were assembled and annealed at 923 K, 973 K, 1073 K, 1173 K, and 1273 K (650 °C, 700 °C, 800 °C, 900 °C, and 1000 °C) for various times. The interdiffusion microstructures and concentration profiles were examined via scanning electron microscopy and electron probe microanalysis, respectively. As the Mo concentration increased from 2 to 26 at. pct, the interdiffusion coefficient decreased, while the activation energy increased. A Kirkendall marker plane was clearly identified in each diffusion couple and utilized to determine intrinsic diffusion coefficients. Uranium intrinsically diffused 5-10 times faster than Mo. Molar excess Gibbs free energy of U-Mo alloy was applied to calculate the thermodynamic factor using ideal, regular, and subregular solution models. Based on the intrinsic diffusion coefficients and thermodynamic factors, Manning's formalism was used to calculate the tracer diffusion coefficients, atomic mobilities, and vacancy wind parameters of U and Mo at the marker composition. The tracer diffusion coefficients and atomic mobilities of U were about five times larger than those of Mo, and the vacancy wind effect increased the intrinsic flux of U by approximately 30 pct.

Thermal properties for the thermal-hydraulics analyses of the BR2 maximum nominal heat flux.

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

Dionne, B.; Kim, Y. S.; Hofman, G. L.

2011-05-23

This memo describes the assumptions and references used in determining the thermal properties for the various materials used in the BR2 HEU (93% enriched in {sup 235}U) to LEU (19.75% enriched in {sup 235}U) conversion feasibility analysis. More specifically, this memo focuses on the materials contained within the pressure vessel (PV), i.e., the materials that are most relevant to the study of impact of the change of fuel from HEU to LEU. This section is regrouping all of the thermal property tables. Section 2 provides a summary of the thermal properties in form of tables while the following sections presentmore » the justification of these values. Section 3 presents a brief background on the approach used to evaluate the thermal properties of the dispersion fuel meat and specific heat capacity. Sections 4 to 7 discuss the material properties for the following materials: (i) aluminum, (ii) dispersion fuel meat (UAlx-Al and U-7Mo-Al), (iii) beryllium, and (iv) stainless steel. Section 8 discusses the impact of irradiation on material properties. Section 9 summarizes the material properties for typical operating temperatures. Appendix A elaborates on how to calculate dispersed phase's volume fraction. Appendix B shows the evolution of the BR2 maximum heat flux with burnup.« less

Thermal Properties for the Thermal-Hydraulics Analyses of the BR2 Maximum Nominal Heat Flux

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

Dionne, B.; Bergeron, A.; Licht, J. R.

2015-02-01

This memo describes the assumptions and references used in determining the thermal properties for the various materials used in the BR2 HEU (93% enriched in 235U) to LEU (19.75% enriched in 235U) conversion feasibility analysis. More specifically, this memo focuses on the materials contained within the pressure vessel (PV), i.e., the materials that are most relevant to the study of impact of the change of fuel from HEU to LEU. Section 2 provides a summary of the thermal properties in the form of tables while the following sections and appendices present the justification of these values. Section 3 presents amore » brief background on the approach used to evaluate the thermal properties of the dispersion fuel meat and specific heat capacity. Sections 4 to 7 discuss the material properties for the following materials: i) aluminum, ii) dispersion fuel meat (UAlx-Al and U-7Mo-Al), iii) beryllium, and iv) stainless steel. Section 8 discusses the impact of irradiation on material properties. Section 9 summarizes the material properties for typical operating temperatures. Appendix A elaborates on how to calculate dispersed phase’s volume fraction. Appendix B provides a revised methodology for determining the thermal conductivity as a function of burnup for HEU and LEU.« less

Modeling Early-Stage Processes of U-10 Wt.%Mo Alloy Using Integrated Computational Materials Engineering Concepts

NASA Astrophysics Data System (ADS)

Wang, Xiaowo; Xu, Zhijie; Soulami, Ayoub; Hu, Xiaohua; Lavender, Curt; Joshi, Vineet

2017-12-01

Low-enriched uranium alloyed with 10 wt.% molybdenum (U-10Mo) has been identified as a promising alternative to high-enriched uranium. Manufacturing U-10Mo alloy involves multiple complex thermomechanical processes that pose challenges for computational modeling. This paper describes the application of integrated computational materials engineering (ICME) concepts to integrate three individual modeling components, viz. homogenization, microstructure-based finite element method for hot rolling, and carbide particle distribution, to simulate the early-stage processes of U-10Mo alloy manufacture. The resulting integrated model enables information to be passed between different model components and leads to improved understanding of the evolution of the microstructure. This ICME approach is then used to predict the variation in the thickness of the Zircaloy-2 barrier as a function of the degree of homogenization and to analyze the carbide distribution, which can affect the recrystallization, hardness, and fracture properties of U-10Mo in subsequent processes.

The structure of premixed particle-cloud flames

NASA Technical Reports Server (NTRS)

Seshadri, K.; Berlad, A. L.; Tangirala, V.

1992-01-01

The structure of premixed flames propagating in combustible systems, containing uniformly distributed volatile fuel particles, in an oxidizing gas mixture, is analyzed. It is presumed that the fuel particles vaporize first to yield a gaseous fuel of known chemical structure, which is subsequently oxidized in the gas phase. The analysis is performed in the asymptotic limit, where the value of the characteristic Zeldovich number, based on the gas-phase oxidation of the gaseous fuel is large, and for values of phi(u) greater than or equal to 1.0, where phi(u) is the equivalence ratio based on the fuel available in the fuel particles. The structure of the flame is presumed to consist of a preheat vaporization zone where the rate of the gas-phase chemical reaction is small, a reaction zone where convection and the rate of vaporization of the fuel particles are small and a convection zone where diffusive terms in the conservation equations are small. For given values phi(u) the analysis yields results for the burning velocity and phi(g) where phi(g) is the effective equivalence ratio in the reaction zone. The analysis shows that even though phi(u) greater than or equal to 1.0, for certain cases the calculated value of phi(g) is less than unity. This prediction is in agreement with experimental observations.

In situ TEM and synchrotron characterization of U–10Mo thin specimen annealed at the fast reactor temperature regime

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

Yun, Di, E-mail: diyun1979@xjtu.edu.cn; Xi'an Jiao Tong University, 28 Xian Ning West Road, Xi'an 710049; Mo, Kun

2015-12-15

U–Mo metallic alloys have been extensively used for the Reduced Enrichment for Research and Test Reactors (RERTR) program, which is now known as the Office of Material Management and Minimization under the Conversion Program. This fuel form has also recently been proposed as fast reactor metallic fuels in the recent DOE Ultra-high Burnup Fast Reactor project. In order to better understand the behavior of U–10Mo fuels within the fast reactor temperature regime, a series of annealing and characterization experiments have been performed. Annealing experiments were performed in situ at the Intermediate Voltage Electron Microscope (IVEM-Tandem) facility at Argonne National Laboratorymore » (ANL). An electro-polished U–10Mo alloy fuel specimen was annealed in situ up to 700 °C. At an elevated temperature of about 540 °C, the U–10Mo specimen underwent a relatively slow microstructure transition. Nano-sized grains were observed to emerge near the surface. At the end temperature of 700 °C, the near-surface microstructure had evolved to a nano-crystalline state. In order to clarify the nature of the observed microstructure, Laue diffraction and powder diffraction experiments were carried out at beam line 34-ID of the Advanced Photon Source (APS) at ANL. Phases present in the as-annealed specimen were identified with both Laue diffraction and powder diffraction techniques. The U–10Mo was found to recrystallize due to thermally-induced recrystallization driven by a high density of pre-existing dislocations. A separate in situ annealing experiment was carried out with a Focused Ion Beam processed (FIB) specimen. A similar microstructure transition occurred at a lower temperature of about 460 °C with a much faster transition rate compared to the electro-polished specimen. - Highlights: • TEM annealing experiments were performed in situ at the IVEM facility up to fast reactor temperature. • At 540 °C, the U-10Mo specimen underwent a slow microstructure transition where nano-sized grains were observed to emerge. • UO{sub 2} phase exists at the thin area of the as-annealed specimen whereas U-10Mo γ phase dominated at the thicker part. • Bcc γ U-10Mo recrystallized to become nano-meter sized crystallites near the specimen surface. • A separateannealing experiment was conducted with a FIB processed specimen where similar transition occurred at a lower temperature of 460 °C with a faster rate.« less

A U-bearing composite waste form for electrochemical processing wastes

NASA Astrophysics Data System (ADS)

Chen, X.; Ebert, W. L.; Indacochea, J. E.

2018-04-01

Metallic/ceramic composite waste forms are being developed to immobilize combined metallic and oxide waste streams generated during electrochemical recycling of used nuclear fuel. Composites were made for corrosion testing by reacting HT9 steel to represent fuel cladding, Zr and Mo to simulate metallic fuel waste, and a mixture of ZrO2, Nd2O3, and UO2 to represent oxide wastes. More than half of the added UO2 was reduced to metal and formed Fe-Zr-U intermetallics and most of the remaining UO2 and all of the Nd2O3 reacted to form zirconates. Fe-Cr-Mo intermetallics were also formed. Microstructure characterization of the intermetallic and ceramic phases that were generated and tests conducted to evaluate their corrosion behaviors indicate composite waste forms can accommodate both metallic and oxidized waste streams in durable host phases.

A U-bearing composite waste form for electrochemical processing wastes

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

Chen, X.; Ebert, W. L.; Indacochea, J. E.

Metallic/ceramic composite waste forms are being developed to immobilize combined metallic and oxide waste streams generated during electrochemical recycling of used nuclear fuel. Composites were made for corrosion testing by reacting HT9 steel to represent fuel cladding, Zr and Mo to simulate metallic fuel waste, and a mixture of ZrO2, Nd2O3, and UO2 to represent oxide wastes. More than half of the added UO2 was reduced to metal and formed Fe-Zr-U intermetallics and most of the remaining UO2 and all of the Nd2O3 reacted to form zirconates. Fe-Cr-Mo intermetallics were also formed. Microstructure characterization of the intermetallic and ceramic phasesmore » that were generated and tests conducted to evaluate their corrosion behaviors indicate composite waste forms can accommodate both metallic and oxidized waste streams in durable host phases. (c) 2018 Elsevier B.V. All rights reserved.« less

Calculation and comparison of xenon and samarium reactivities of the HEU, LEU core in the low power research reactor.

PubMed

Dawahra, S; Khattab, K; Saba, G

2015-07-01

Comparative studies for the conversion of the fuel from HEU to LEU in the Miniature Neutron Source Reactor (MNSR) have been performed using the MCNP4C and GETERA codes. The precise calculations of (135)Xe and (149)Sm concentrations and reactivities were carried out and compared during the MNSR operation time and after shutdown for the existing HEU fuel (UAl4-Al, 90% enriched) and the potential LEU fuels (U3Si2-Al, U3Si-Al, U9Mo-Al, 19.75% enriched and UO2, 12.6% enriched) in this paper using the MCNP4C and GETERA codes. It was found that the (135)Xe and (149)Sm reactivities did not reach their equilibrium reactivities during the daily operating time of the reactor. The (149)Sm reactivities could be neglected compared to (135)Xe reactivities during the reactor operating time and after shutdown. The calculations for the UAl4-Al produced the highest (135)Xe reactivity in all the studied fuel group during the reactor operation (0.39 mk) and after the reactor shutdown (0.735 mk), It followed by U3Si-Al (0.34 mk, 0.653 mk), U3Si2-Al (0.33 mk, 0.634 mk), U9Mo-Al (0.3 mk, 0.568 mk) and UO2 (0.24 mk, 0.448 mk) fuels, respectively. Finally, the results showed that the UO2 was the best candidate for fuel conversion to LEU in the MNSR since it gave the lowest (135)Xe reactivity during the reactor operation and after shutdown. Copyright © 2015 Elsevier Ltd. All rights reserved.

Production of LEU Fully Ceramic Microencapsulated Fuel for Irradiation Testing

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

Terrani, Kurt A; Kiggans Jr, James O; McMurray, Jake W

2016-01-01

Fully Ceramic Microencapsulated (FCM) fuel consists of tristructural isotropic (TRISO) fuel particles embedded inside a SiC matrix. This fuel inherently possesses multiple barriers to fission product release, namely the various coating layers in the TRISO fuel particle as well as the dense SiC matrix that hosts these particles. This coupled with the excellent oxidation resistance of the SiC matrix and the SiC coating layer in the TRISO particle designate this concept as an accident tolerant fuel (ATF). The FCM fuel takes advantage of uranium nitride kernels instead of oxide or oxide-carbide kernels used in high temperature gas reactors to enhancemore » heavy metal loading in the highly moderated LWRs. Production of these kernels with appropriate density, coating layer development to produce UN TRISO particles, and consolidation of these particles inside a SiC matrix have been codified thanks to significant R&D supported by US DOE Fuel Cycle R&D program. Also, surrogate FCM pellets (pellets with zirconia instead of uranium-bearing kernels) have been neutron irradiated and the stability of the matrix and coating layer under LWR irradiation conditions have been established. Currently the focus is on production of LEU (7.3% U-235 enrichment) FCM pellets to be utilized for irradiation testing. The irradiation is planned at INL s Advanced Test Reactor (ATR). This is a critical step in development of this fuel concept to establish the ability of this fuel to retain fission products under prototypical irradiation conditions.« less

US RERTR FUEL DEVELOPMENT POST IRRADIATION EXAMINATION RESULTS

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

A. B. Robinson; D. M. Wachs; D. E. Burkes

2008-10-01

Post irradiation examinations of irradiated RERTR plate type fuel at the Idaho National Laboratory have led to in depth characterization of fuel behavior and performance. Both destructive and non-destructive examination capabilities at the Hot Fuels Examination Facility (HFEF) as well as recent results obtained are discussed herein. New equipment as well as more advanced techniques are also being developed to further advance the investigation into the performance of the high density U-Mo fuel.

Ethanol Blend Effects On Direct Injection Spark-Ignition Gasoline Vehicle Particulate Matter Emissions

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

Storey, John Morse; Lewis Sr, Samuel Arthur; Barone, Teresa L

2010-01-01

Direct injection spark-ignition (DISI) gasoline engines can offer better fuel economy and higher performance over their port fuel-injected counterparts, and are now appearing increasingly in more U.S. vehicles. Small displacement, turbocharged DISI engines are likely to be used in lieu of large displacement engines, particularly in light-duty trucks and sport utility vehicles, to meet fuel economy standards for 2016. In addition to changes in gasoline engine technology, fuel composition may increase in ethanol content beyond the 10% allowed by current law due to the Renewable Fuels Standard passed as part of the 2007 Energy Independence and Security Act (EISA). Inmore » this study, we present the results of an emissions analysis of a U.S.-legal stoichiometric, turbocharged DISI vehicle, operating on ethanol blends, with an emphasis on detailed particulate matter (PM) characterization. Gaseous species, particle mass, and particle number concentration emissions were measured for the Federal Test Procedure urban driving cycle (FTP 75) and the more aggressive US06 cycle. Particle number-size distributions and organic to elemental carbon ratios (OC/EC) were measured for 30 MPH and 80 MPH steady-state operation. In addition, particle number concentration was measured during wide open throttle accelerations (WOTs) and gradual accelerations representative of the FTP 75. For the gaseous species and particle mass measurements, dilution was carried out using a full flow constant volume sampling system (CVS). For the particle number concentration and size distribution measurements, a micro-tunnel dilution system was employed. The vehicles were fueled by a standard test gasoline and 10% (E10) and 20% (E20) ethanol blends from the same supplier. The particle mass emissions were approximately 3 and 7 mg/mile for the FTP75 and US06, respectively, with lower emissions for the ethanol blends. During steady-state operation, the geometric mean diameter of the particle-number size distribution remained approximately the same (50 nm) but the particle number concentration decreased with increasing ethanol content in the fuel. In addition, increasing ethanol content significantly reduced the number concentration of 50 and 100 nm particles during gradual and WOT accelerations.« less

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

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

Rest, J.; Hofman, G.L.

1997-12-01

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

FLOW TESTING AND ANALYSIS OF THE FSP-1 EXPERIMENT

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

Hawkes, Grant L.; Jones, Warren F.; Marcum, Wade

The U.S. High Performance Research Reactor Conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel. The Full Scale Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Flow testing experimentation and hydraulic analysis have been performed on the FSP-1 experiment to be irradiated inmore » the ATR at the Idaho National Laboratory (INL). A flow test experiment mockup of the FSP-1 experiment was completed at Oregon State University. Results of several flow test experiments are compared with analyses. This paper reports and shows hydraulic analyses are nearly identical to the flow test results. A water velocity of 14.0 meters per second is targeted between the fuel plates. Comparisons between FSP-1 measurements and this target will be discussed. This flow rate dominates the flow characteristics of the experiment and model. Separate branch flows have minimal effect on the overall experiment. A square flow orifice was placed to control the flowrate through the experiment. Four different orifices were tested. A flow versus delta P curve for each orifice is reported herein. Fuel plates with depleted uranium in the fuel meat zone were used in one of the flow tests. This test was performed to evaluate flow test vibration with actual fuel meat densities and reported herein. Fuel plate deformation tests were also performed and reported.« less

Investigation of a Complex Technique of Smoke Particle Deposition on Scavengers

DTIC Science & Technology

1987-03-01

Code) 7b. ADDRESS (CiJyjoj~ e and ZIP Co* e ) Rola, O 6401P.O. BoRolla MO 6’~OlResearch Triangle Park, NC 27709 Sa. NAME OF FUNOINGiSPONSO.IING O b. OFFICE...TITLE (Inld~ e Secunty Clew V11111’ non) . IN Investigation of a Co~mplex Tecbnique of Smoke Particle Deposition on Scavengers 12 PERSONAL AUTHOR(S...was authorized under Contract No. DAAK-1l-83-K-0007. This work was started in ’u,; e 1983 and completed in August 1986. The use of trade names or

Coupled Mo-U abundances and isotopes in a small marine euxinic basin: Constraints on processes in euxinic basins

NASA Astrophysics Data System (ADS)

Bura-Nakić, Elvira; Andersen, Morten B.; Archer, Corey; de Souza, Gregory F.; Marguš, Marija; Vance, Derek

2018-02-01

Sedimentary molybdenum (Mo) and uranium (U) abundances, as well as their isotope systematics, are used to reconstruct the evolution of the oxygenation state of the surface Earth from the geological record. Their utility in this endeavour must be underpinned by a thorough understanding of their behaviour in modern settings. In this study, Mo-U concentrations and their isotope compositions were measured in the water column, sinking particles, sediments and pore waters of the marine euxinic Lake Rogoznica (Adriatic Sea, Croatia) over a two year period, with the aim of shedding light on the specific processes that control Mo-U accumulation and isotope fractionations in anoxic sediment. Lake Rogoznica is a 15 m deep stratified sea-lake that is anoxic and euxinic at depth. The deep euxinic part of the lake generally shows Mo depletions consistent with near-quantitative Mo removal and uptake into sediments, with Mo isotope compositions close to the oceanic composition. The data also, however, show evidence for periodic additions of isotopically light Mo to the lake waters, possibly released from authigenic precipitates formed in the upper oxic layer and subsequently processed through the euxinic layer. The data also show evidence for a small isotopic offset (∼0.3‰ on 98Mo/95Mo) between particulate and dissolved Mo, even at highest sulfide concentrations, suggesting minor Mo isotope fractionation during uptake into euxinic sediments. Uranium concentrations decrease towards the bottom of the lake, where it also becomes isotopically lighter. The U systematics in the lake show clear evidence for a dominant U removal mechanism via diffusion into, and precipitation in, euxinic sediments, though the diffusion profile is mixed away under conditions of increased density stratification between an upper oxic and lower anoxic layer. The U diffusion-driven precipitation is best described with an effective 238U/235U fractionation of +0.6‰, in line with other studied euxinic basins. Combining the Mo and U systematics in Lake Rogoznica and other euxinic basins, it is apparent that the two different uptake mechanisms of U and Mo can lead to spatially and temporally variable Mo/U and Mo-U isotope systematics that depend on the rate of water renewal versus removal to sediment, the sulfide concentration, and the geometry of the basin. This study further emphasises the potential of combining multiple observations, from Mo-U enrichment and isotope systematics, for disentangling the various processes via which redox conditions control the chemistry of modern and ancient sediments.

Synthesis and Characterization of Molybdenum Based Colloidal Particles.

PubMed

Moreno; Vidoni; Ovalles; Chaudret; Urbina; Krentzein

1998-11-15

The synthesis and characterization of molybdenum colloidal particles were evaluated using thermal and sonochemical methods and starting from different metal precursors, Mo(CO)6 and (NH4)2MoS4. The products were characterized by elemental analysis, spectroscopic (UV, FTIR), and surface analysis (XPS) techniques, as well as by transmission electron microscopy (TEM) for determining the particle sizes. Using Mo(CO)6 as metal source, particle sizes with an average diameter of 1.5 nm can be obtained using tert-amyl alcohol as solvent and tetrahydrothiophene as sulfurating ligand. The characterization of these particles showed that they are composed of molybdenum oxide MoO3. Using (NH4)2MoS4 as metal precursor, particles with average diameters of 4.7 and 2.5 nm were synthesized using thermal and sonochemical methods, respectively. The characterization of these particles showed them to be composed of molybdenum sulfide, MoS2. The sonochemical method proved to be the fastest and most convenient synthetic pathway of obtaining small colloidal particles at low temperatures and with control of the average size. Copyright 1998 Academic Press.

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

Short communication on Kinetics of grain growth and particle pinning in U-10 wt.% Mo

NASA Astrophysics Data System (ADS)

Frazier, William E.; Hu, Shenyang; Overman, Nicole; Lavender, Curt; Joshi, Vineet V.

2018-01-01

The alloy U-10 wt% Mo was annealed at temperatures ranging from 700 °C to 900 °C for periods lasting up to 24 h. Annealed microstructures were examined using Electron Backscattered Diffraction (EBSD) to obtain average grain sizes and grain size distributions. From the temporal evolution of the average grain size, the activation energy of grain growth was determined to be 172.4 ± 0.961 kJ/mol. Grain growth over the annealing period stagnated after a period of 1-4 h. This stagnation is apparently caused by the pinning effect of second-phase particles in the materials. Back-scattered electron imaging (BSE) was used to confirm that these particles do not appreciably coarsen or dissolve during annealing at the aforementioned temperatures.

Progress on RERTR activities in Argentina

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

Balart, S.; Calzetta, O.; Cristini, P.

2008-07-15

Since last RERTR meeting, several tasks involving RERTR activities continued deploying in Argentina: through an agreement between CNEA and US-DoE final steps in the RA-6 reactor core conversion from HEU to LEU are taking place; by means of a return campaign of 42 US origin SNF in the frame of the US-SNF FRR program; an effective minimization of HEU inventory is close to be accomplished; development of a LEU dispersed U-Mo fuel prototype, to be irradiated in a high flux reactor in the frame of the ARG/4/092 IAEA's Technical Cooperation project is progressing; very high density monolithic U-Mo miniplates andmore » plates using MEU and LEU fuel with Zry-4 cladding were developed to be irradiated as a part of the RERTR program irradiation experiment; atomistic modeling prediction (BFS techniques and first principles) enabled to find some trends on the interaction phases; diffusion couples tests under X-ray synchrotron analysis allowed the characterization of several phases involving U-Mo(-Zr) / Al(-Si); finally CNEA continued spreading high quality LEU technology for fission RI production by means of agreements with different producers interested on HEU-LEU conversion. (author)« less

Exhaust particle characterization for lean and stoichiometric DI vehicles operating on ethanol-gasoline blends

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

Storey, John Morse; Barone, Teresa L; Thomas, John F

2012-01-01

Gasoline direct injection (GDI) engines can offer better fuel economy and higher performance over their port fuel-injected (PFI) counterparts, and are now appearing in increasingly more U.S. and European vehicles. Small displacement, turbocharged GDI engines are replacing large displacement engines, particularly in light-duty trucks and sport utility vehicles, in order for manufacturers to meet the U.S. fuel economy standards for 2016. Furthermore, lean-burn GDI engines can offer even higher fuel economy than stoichiometric GDI engines and have overcome challenges associated with cost-effective aftertreatment for NOx control. Along with changes in gasoline engine technology, fuel composition may increase in ethanol contentmore » beyond the current 10% due to the recent EPA waiver allowing 15% ethanol. In addition, the Renewable Fuels Standard passed as part of the 2007 Energy Independence and Security Act (EISA) mandates the use of biofuels in upcoming years. GDI engines are of environmental concern due to their high particulate matter (PM) emissions relative to port-fuel injected (PFI) gasoline vehicles; widespread market penetration of GDI vehicles may result in additional PM from mobile sources at a time when the diesel contribution is declining. In this study, we characterized particulate emissions from a European certified lean-burn GDI vehicle operating on ethanol-gasoline blends. Particle mass and particle number concentration emissions were measured for the Federal Test Procedure urban driving cycle (FTP 75) and the more aggressive US06 driving cycle. Particle number-size distributions and organic to elemental carbon ratios (OC/EC) were measured for 30 MPH and 80 MPH steady-state operation. In addition, particle number concentration was measured during wide open throttle accelerations (WOTs) and gradual accelerations representative of the FTP 75. Fuels included certification gasoline and 10% (E10) and 20% (E20) ethanol blends from the same supplier. The particle mass emissions were approximately 3 and 7 mg/mile for the FTP75 and US06, respectively, with lower emissions for the ethanol blends. The data are compared to a previous study on a U.S.-legal stoichiometric GDI vehicle operating on the same ethanol blends. The lean-burn GDI vehicle emitted a higher number of particles, but had an overall smaller average size. Particle number per mile decreased with increasing ethanol content for the transient tests. For the 30 and 80 mph tests, particle number concentration decreased with increasing ethanol content, although the shape of the particle size distribution remained the same. Engine-out OC/EC ratios were highest for the stoichiometric GDI vehicle with E20, but tailpipe OC/EC ratios were similar for all vehicles.« less

RERTR-8 Irradiation Summary Report

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

D. M. Perez; M. A. Lillo; G. S. Chang

2011-12-01

The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-8, was designed to test monolithic mini-fuel plates fabricated via hot isostatic pressing (HIP), the effect of molybdenum (Mo) content on the monolithic fuel behavior, and the efficiency of ternary additions to dispersion fuel particles on the interaction layer behavior at higher burnup. The following report summarizes the life of the RERTR-8 experiment through end of irradiation, including as-run neutronic analysis, thermal analysis and hydraulic testing results.

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

Liang, Linyun; Mei, Zhi-Gang; Yacout, Abdellatif M.

We have developed a mesoscale phase-field model for studying the effect of recrystallization on the gas-bubble-driven swelling in irradiated U-Mo alloy fuel. The model can simulate the microstructural evolution of the intergranular gas bubbles on the grain boundaries as well as the recrystallization process. Our simulation results show that the intergranular gas-bubble-induced fuel swelling exhibits two stages: slow swelling kinetics before recrystallization and rapid swelling kinetics with recrystallization. We observe that the recrystallization can significantly expedite the formation and growth of gas bubbles at high fission densities. The reason is that the recrystallization process increases the nucleation probability of gasmore » bubbles and reduces the diffusion time of fission gases from grain interior to grain boundaries by increasing the grain boundary area and decreasing the diffusion distance. The simulated gas bubble shape, size distribution, and density on the grain boundaries are consistent with experimental measurements. We investigate the effect of the recrystallization on the gas-bubble-driven fuel swelling in UMo through varying the initial grain size and grain aspect ratio. We conclude that the initial microstructure of fuel, such as grain size and grain aspect ratio, can be used to effectively control the recrystallization and therefore reduce the swelling in U-Mo fuel.« less

GDGT and alkenone flux in the northern Gulf of Mexico: Implications for the TEX86 and UK'37 paleothermometers

NASA Astrophysics Data System (ADS)

Richey, Julie N.; Tierney, Jessica E.

2016-12-01

The TEX86 and U37K' molecular biomarker proxies have been broadly applied in downcore marine sediments to reconstruct past sea surface temperature (SST). Although both TEX86 and U37K' have been interpreted as proxies for mean annual SST throughout the global ocean, regional studies of glycerol dibiphytanyl glycerol tetraethers (GDGTs) and alkenones in sinking particles are required to understand the influence of seasonality, depth distribution, and diagenesis on downcore variability. We measure GDGT and alkenone flux, as well as the TEX86 and U37K' indices in a 4 year sediment trap time series (2010-2014) in the northern Gulf of Mexico (nGoM), and compare these data with core-top sediments at the same location. GDGT and alkenone fluxes do not show a consistent seasonal cycle; however, the largest flux peaks for both occurs in winter. U37K' covaries with SST over the 4 year sampling interval, but the U37K'-SST relationship in this data set implies a smaller slope or nonlinearity at high temperatures when compared with existing calibrations. Furthermore, the flux-weighted U37K' value from sinking particles is significantly lower than that of underlying core-top sediments, suggesting preferential diagenetic loss of the tri-unsaturated alkenone in sediments. TEX86 does not covary with SST, suggesting production in the subsurface upper water column. The flux-weighted mean TEX86 matches that of core-top sediments, confirming that TEX86 in the nGoM reflects local planktonic production rather than allochthonous or in situ sedimentary production. We explore potential sources of uncertainty in both proxies in the nGoM but demonstrate that they show nearly identical trends in twentieth century SST, despite these factors.

An Integrated Computational and Experimental Approach Toward the Design of Materials for Fuel Cell Systems

DTIC Science & Technology

2012-10-01

13 Based on the limited work done, the best reported ORR chalcogenide electrocatalysts for PEMFC applications can be ranked as follows: MoRuSe... PEMFC catalysts is the durability of the catalyst particles. Particle size distribution tends to shift towards larger particles during the...the design of new materials for applications in PEMFCs . Reference: A more detailed treatment of the topics of this section, Experimental Target 11

Study of a generalized birks formula for the scintillation response of a CaMoO4 crystal

NASA Astrophysics Data System (ADS)

Lee, J. Y.; Kim, H. J.; Kang, Sang Jun; Lee, M. H.

2017-12-01

We have investigated the scintillation characteristics of CaMoO4 (CMO) crystals by using a gamma source and various internal alpha sources. A 137Cs source with 662-keV gamma-rays was used for the gamma-quanta light yield calibration. Internal radioactive contaminations provided alpha particles with different energies from 5.41 to 7.88 MeV. We developed a C++ program based on the ROOT package for the fitting of parameters in a generalized Birks semi-empirical formula by combining the experimental and the simulation data. Results for the fitted Birks parameters are k b1 = 3.3 × 10 -3 (g/MeVcm2) for the 1st parameter and k b2 = 7.9 × 10 -5 (g/MeVcm2)2 for the 2nd parameter. The χ2/n.d.f. (Number of Degree of Freedom) is calculated as 0.1/4. We were able to estimate the 238U and 234U contaminations in a CMO crystal by using the generalized Birks semi-empirical formula.

Design and Status of RERTR Irradiation Tests in the Advanced Test Reactor

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

Daniel M. Wachs; Richard G. Ambrosek; Gray Chang

2006-10-01

Irradiation testing of U-Mo based fuels is the central component of the Reduced Enrichment for Research and Test Reactors (RERTR) program fuel qualification plan. Several RERTR tests have recently been completed or are planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory in Idaho Falls, ID. Four mini-plate experiments in various stages of completion are described in detail, including the irradiation test design, objectives, and irradiation conditions. Observations made during and after the in-reactor RERTR-7A experiment breach are summarized. The irradiation experiment design and planned irradiation conditions for full-size plate test are described. Progressmore » toward element testing will be reviewed.« less

Durability test on irradiated rock-like oxide fuels

NASA Astrophysics Data System (ADS)

Kuramoto, K.; Nitani, N.; Yamashita, T.

2003-06-01

For a profitable use of Pu, Japan Atomic Energy Research Institute has been promoting researches for once-through type fuels. The strategy consists of stable rock-like oxide fuel fabrication in conventional fuel facilities followed by almost complete Pu burning in LWR and disposal of chemically stable spent fuel without further processing. Because leach rates of hazardous nuclides, such as TRU and β-emitters, that have long half-lives, are very important for the evaluation of geological safety, leaching tests in deionized water at 363 K were performed with reference to the MCC-1 method. Five irradiated fuel pellets, a single phase fuel of a yttria-stabilized zirconia (YSZ) containing UO 2 (U-YSZ), two fuels of U-YSZ particle dispersed in MgAl 2O 4 (SPI) or Al 2O 3 (COR) matrix, two homogeneous-blended fuels of U-YSZ and SPI or COR powders, were submitted to the tests. Stainless steel containers with Au coating and ethylene propylene diene monomer were used as leaching vessels and packing, respectively. The evaluated normalized leach rates of Zr, U and Pu were obviously lower than those of the other important elements and nuclides. Americium, Np and especially Y showed unexpectedly high evaluated normalized leach rates. The volatile elements, Cs and I, showed enhanced leaching within particle-dispersed type fuels because of crack formation around the particle.

Performance evaluation of bimodal thermite composites : nano- vs miron-scale particles

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

Moore, K. M.; Pantoya, M.; Son, S. F.

2004-01-01

In recent years many studies of metastable interstitial composites (MIC) have shown vast combustion improvements over traditional thermite materials. The main difference between these two materials is the size of the fuel particles in the mixture. Decreasing the fuel size from the micron to nanometer range significantly increases the combustion wave speed and ignition sensitivity. Little is known, however, about the critical level of nano-sized fuel particles needed to enhance the performance of the traditional thermite. Ignition sensitivity experiments were performed using Al/MoO{sub 3} pellets at a theoretical maximum density of 50% (2 g/cm{sup 3}). The Al fuel particles weremore » prepared as bi-modal size distributions with micron (i.e., 4 and 20 {micro}m diameter) and nano-scale Al particles. The micron-scale Al was replaced in 10% increments by 80 nm Al particles until the fuel was 100% 80 nm Al. These bi-modal distributions allow the unique characteristics of nano-scale materials to be better understood. The pellets were ignited using a 50-W CO{sub 2} laser. High speed imaging diagnostics were used to measure ignition delay times, and micro-thermocouples were used to measure ignition temperatures. Combustion wave speeds were also examined.« less

Thermite combustion enhancement resulting from biomodal luminum distribution

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

Moore, K. M.; Pantoya, M.; Son, S. F.

2004-01-01

In recent years many studies that incorporated nano-scale or ultrafine aluminum (Al) as part of an energetic formulation and demonstrated significant performance enhancement. Decreasing the fuel particle size from the micron to nanometer range alters the material's chemical and thermal-physical properties. The result is increased particle reactivity that translates to an increase in the combustion wave speed and ignition sensitivity. Little is known, however, about the critical level of nano-sized fuel particles needed to enhance the performance of the energetic composite. Ignition sensitivity and combustion wave speed experiments were performed using a thermite composite of Al and MoO{sub 3} pressedmore » to a theoretical maximum density of 50% (2 g/cm{sup 3}). A bimodal Al particle size distribution was prepared using 4 or 20 {mu}m Al fuel particles that were replaced in 10% increments by 80 nm Al particles until the fuel was 100% 80 nm Al. These bimodal distributions allow the unique characteristics of nano-scale materials to be better understood. The pellets were ignited using a 50W CO{sub 2} laser. High speed imaging diagnostics were used to measure the ignition delay time and combustion wave speed.« less

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

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

Rest, J.; Hofman, G.L.

1999-04-01

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

Engineering Ni-Mo-S Nanoparticles for Hydrodesulfurization.

PubMed

Bodin, Anders; Christoffersen, Ann-Louise N; Elkjær, Christian F; Brorson, Michael; Kibsgaard, Jakob; Helveg, Stig; Chorkendorff, Ib

2018-06-13

Nanoparticle engineering for catalytic applications requires both a synthesis technique for the production of well-defined nanoparticles and measurements of their catalytic performance. In this paper, we present a new approach to rationally engineering highly active Ni-Mo-S nanoparticle catalysts for hydrodesulfurization (HDS), i.e., the removal of sulfur from fossil fuels. Nanoparticle catalysts are synthesized by the sputtering of a Mo 75 Ni 25 metal target in a reactive atmosphere of Ar and H 2 S followed by the gas aggregation of the sputtered material into nanoparticles. The nanoparticles are filtered by a quadrupole mass filter and subsequently deposited on a planar substrate, such as a grid for electron microscopy or a microreactor. By varying the mass of the deposited nanoparticles, it is demonstrated that the Ni-Mo-S nanoparticles can be tuned into fullerene-like particles, flat-lying platelets, and upright-oriented platelets. The nanoparticle morphologies provide different abundances of Ni-Mo-S edge sites, which are commonly considered the catalytically important sites. Using a microreactor system, we assess the catalytic activity of the Ni-Mo-S nanoparticles for the HDS of dibenzothiophene. The measurements show that platelets are twice as active as the fullerene-like particles, demonstrating that the Ni-Mo-S edges are more active than basal planes for the HDS. Furthermore, the upright-standing orientation of platelets show an activity that is six times higher than the fullerene-like particles, demonstrating the importance of the edge site number and accessibility to reducing, e.g., sterical hindrance for the reacting molecules.

Structural and thermodynamic study of dicesium molybdate Cs2Mo2O7: Implications for fast neutron reactors

NASA Astrophysics Data System (ADS)

Smith, A. L.; Kauric, G.; van Eijck, L.; Goubitz, K.; Wallez, G.; Griveau, J.-C.; Colineau, E.; Clavier, N.; Konings, R. J. M.

2017-09-01

The structure of α-Cs2Mo2O7 (monoclinic in space group P21 / c), which can form during irradiation in fast breeder reactors in the space between nuclear fuel and cladding, has been refined in this work at room temperature from neutron diffraction data. Furthermore, the compounds' thermal expansion and polymorphism have been investigated using high temperature X-ray diffraction combined with high temperature Raman spectroscopy. A phase transition has been observed at Ttr(α → β)=(621.9±0.8) K using Differential Scanning Calorimetry, and the structure of the β-Cs2Mo2O7 phase, orthorhombic in space group Pbcm, has been solved ab initio from the high temperature X-ray diffraction data. Furthermore, the low temperature heat capacity of α-Cs2Mo2O7 has been measured in the temperature range T=(1.9-313.2) K using a Quantum Design PPMS (Physical Property Measurement System) calorimeter. The heat capacity and entropy values at T=298.15 K have been derived as Cp,mo (Cs2Mo2O7 , cr , 298.15 K) = (211.9 ± 2.1) J K-1mol-1 and Smo (Cs2Mo2O7 , cr , 298.15 K) = (317.4 ± 4.3) J K-1mol-1 . When combined with the enthalpy of formation reported in the literature, these data yield standard entropy and Gibbs energy of formation as Δf Smo (Cs2Mo2O7 , cr , 298.15 K) = - (628.2 ± 4.4) J K-1mol-1 and Δf Gmo (Cs2Mo2O7 , cr , 298.15 K) = - (2115.1 ± 2.5) kJmol-1 . Finally, the cesium partial pressure expected in the gap between fuel and cladding following the disproportionation reaction 2Cs2MoO4=Cs2Mo2O7+2Cs(g)+ 1/2 O2(g) has been calculated from the newly determined thermodynamic functions.

Atomic scale modelling of hexagonal structured metallic fission product alloys

PubMed Central

Middleburgh, S. C.; King, D. M.; Lumpkin, G. R.

2015-01-01

Noble metal particles in the Mo-Pd-Rh-Ru-Tc system have been simulated on the atomic scale using density functional theory techniques for the first time. The composition and behaviour of the epsilon phases are consistent with high-entropy alloys (or multi-principal component alloys)—making the epsilon phase the only hexagonally close packed high-entropy alloy currently described. Configurational entropy effects were considered to predict the stability of the alloys with increasing temperatures. The variation of Mo content was modelled to understand the change in alloy structure and behaviour with fuel burnup (Mo molar content decreases in these alloys as burnup increases). The predicted structures compare extremely well with experimentally ascertained values. Vacancy formation energies and the behaviour of extrinsic defects (including iodine and xenon) in the epsilon phase were also investigated to further understand the impact that the metallic precipitates have on fuel performance. PMID:26064629

HEAT TREATED U-Mo ALLOY

DOEpatents

McGeary, R.K.; Justusson, W.M.

1960-02-23

A reactor fuel element comprising a gamma-phase alloy consisting of 11 to 16 wt.% of molyhdenum and the balance uranium, annealed between 350 and 525 deg C and quenched to preserve the gamma phase, is reported.

SHINE and Mini-SHINE Column Designs for Recovery of Mo from 140 g-U/L Uranyl Sulfate

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

Stepinski, Dominique C.; Vandegrift, George F.

Argonne is assisting SHINE Medical Technologies (SHINE) in their efforts to develop an accelerator-driven process that utilizes a uranyl-sulfate solution for the production of fission Mo-99. In an effort to design a Mo-recovery system for the SHINE project using low-enriched uranium (LEU), we conducted batch, breakthrough, and pulse tests to determine the Mo isotherm, mass-transfer zone (MTZ), and system parameters for a 130 g-U/L uranyl sulfate solution at pH 1 and 80°C, as described previously. The VERSE program was utilized to calculate the MTZ under various loading times and velocities. The results were then used to design Mo separation andmore » recovery columns employing a pure titania sorbent (110-μm particles, S110, and 60 Å pore size). The plant-scale column designs assume Mo will be separated from 271 L of a 141 g-U/L uranyl sulfate solution, pH 1, containing 0.0023 mM Mo. The VERSE-designed recovery systems have been tested and verified in laboratory-scale experiments, and this approach was found to be very successful.« less

Energy recycling by co-combustion of coal and recovered paint solids from automobile paint operations

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

Achariya Suriyawong; Rogan Magee; Ken Peebles

2009-05-15

This paper presents the results of an experimental study of particulate emission and the fate of 13 trace elements (arsenic (As), barium (Ba), cadmium (Cd), chromium (Cr), copper (Cu), cobalt (Co), manganese (Mn), molybdenum (Mo), nickel (Ni), lead (Pb), mercury (Hg), vanadium (V), and zinc (Zn)) during combustion tests of recovered paint solids (RPS) and coal. The emissions from combustions of coal or RPS alone were compared with those of co-combustion of RPS with subbituminous coal. The distribution/partitioning of these toxic elements between a coarse-mode ash (particle diameter (d{sub p}) > 0.5 {mu}m), a submicrometer-mode ash (d{sub p} < 0.5more » {mu}m), and flue gases was also evaluated. Submicrometer particles generated by combustion of RPS alone were lower in concentration and smaller in size than that from combustion of coal. However, co-combustion of RPS and coal increased the formation of submicrometer-sized particles because of the higher reducing environment in the vicinity of burning particles and the higher volatile chlorine species. Hg was completely volatilized in all cases; however, the fraction in the oxidized state increased with co-combustion. Most trace elements, except Zn, were retained in ash during combustion of RPS alone. Mo was mostly retained in all samples. The behavior of elements, except Mn and Mo, varied depending on the fuel samples. As, Ba, Cr, Co, Cu, and Pb were vaporized to a greater extent from cocombustion of RPS and coal than from combustion of either fuel. Evidence of the enrichment of certain toxic elements in submicrometer particles has also been observed for As, Cd, Cr, Cu, and Ni during co-combustion. 27 refs., 6 figs., 5 tabs.« less

Fuel Performance Experiments and Modeling: Fission Gas Bubble Nucleation and Growth in Alloy Nuclear Fuels

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

McDeavitt, Sean; Shao, Lin; Tsvetkov, Pavel

2014-04-07

Advanced fast reactor systems being developed under the DOE's Advanced Fuel Cycle Initiative are designed to destroy TRU isotopes generated in existing and future nuclear energy systems. Over the past 40 years, multiple experiments and demonstrations have been completed using U-Zr, U-Pu-Zr, U-Mo and other metal alloys. As a result, multiple empirical and semi-empirical relationships have been established to develop empirical performance modeling codes. Many mechanistic questions about fission as mobility, bubble coalescience, and gas release have been answered through industrial experience, research, and empirical understanding. The advent of modern computational materials science, however, opens new doors of development suchmore » that physics-based multi-scale models may be developed to enable a new generation of predictive fuel performance codes that are not limited by empiricism.« less

Growth of the interaction layer around fuel particles in dispersion fuel

NASA Astrophysics Data System (ADS)

Olander, D.

2009-01-01

Corrosion of uranium particles in dispersion fuel by the aluminum matrix produces interaction layers (an intermetallic-compound corrosion product) around the shrinking fuel spheres. The rate of this process was modeled as series resistances due to Al diffusion through the interaction layer and reaction of aluminum with uranium in the fuel particle to produce UAl x. The overall kinetics are governed by the relative rates of these two steps, the slowest of which is reaction at the interface between Al in the interaction layer and U in the fuel particle. The substantial volume change as uranium is transferred from the fuel to the interaction layer was accounted for. The model was compared to literature data on in-reactor growth of the interaction layer and the Al/U gradient in this layer, the latter measured in ex-reactor experiments. The rate constant of the Al-U interface reaction and the diffusivity of Al in the interaction layer were obtained from this fitting procedure. The second feature of the corrosion process is the transfer of fission products from the fuel particle to the interaction layer due to the reaction. It is commonly assumed that the observed swelling of irradiated fuel elements of this type is due to release of fission gas in the interaction layer to form large bubbles. This hypothesis was tested by using the model to compute the quantity of fission gas available from this source and comparing the pressure of the resulting gas with the observed swelling of fuel plates. It was determined that the gas pressure so generated is too small to account for the observed delamination of the fuel.

Dissolution of Used Nuclear Fuel Using a TBP/N-Paraffin Solvent

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

Rudisill, T. S.; Shehee, T. C.; Jones, D. H.

2017-10-02

The dissolution of unirradiated used nuclear fuel (UNF) pellets pretreated for tritium removal was demonstrated using a tributly phosphate (TBP) solvent. Dissolution of pretreated fuel in TBP could potentially combine dissolution with two cycle of solvent extraction required for separating the actinides and lanthanides from other fission products. Dissolutions were performed using UNF surrogates prepared from both uranyl nitrate and uranium trioxide produced from the pretreatment process by adding selected actinide and stable fission product elements. In laboratory-scale experiments, the U dissolution efficiency ranged from 80-99+% for both the nitrate and oxide surrogate fuels. On average, 80% of the Pumore » and 50% of the Np and Am in the nitrate surrogate dissolved; however, little of the transuranic elements dissolved in the oxide form. The majority of the 3+ lanthanide elements dissolved. Only small amounts of Sr (0-1.6%) and Mo (0.1-1.7%) and essentially no Cs, Ru, Zr, or Pd dissolved.« less

Premixed flame propagation in combustible particle cloud mixtures

NASA Technical Reports Server (NTRS)

Seshadri, K.; Yang, B.

1993-01-01

The structures of premixed flames propagating in combustible systems, containing uniformly distributed volatile fuel particles, in an oxidizing gas mixtures is analyzed. The experimental results show that steady flame propagation occurs even if the initial equivalence ratio of the combustible mixture based on the gaseous fuel available in the particles, phi(u) is substantially larger than unity. A model is developed to explain these experimental observations. In the model it is presumed that the fuel particles vaporize first to yield a gaseous fuel of known chemical composition which then reacts with oxygen in a one-step overall process. It is shown that the interplay of vaporization kinetics and oxidation process, can result in steady flame propagation in combustible mixtures where the value of phi(u) is substantially larger than unity. This prediction is in agreement with experimental observations.

Electron Microscopic Examination of Irradiated TRISO Coated Particles of Compact 6-3-2 of AGR-1 Experiment

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

Van Rooyen, Isabella Johanna; Demkowicz, Paul Andrew; Riesterer, Jessica Lori

2012-12-01

The electron microscopic examination of selected irradiated TRISO coated particles of the AGR-1 experiment of fuel compact 6-3-2 are presented in this report. Compact 6-3-2 refers to the compact in Capsule 6 at level 3 of Stack 2. The fuel used in capsule 6 compacts, are called the “baseline” fuel as it is fabricated with refined coating process conditions used to fabricate historic German fuel, because of its excellent irradiation performance with UO2 kernels. The AGR-1 fuel is however made of low-enriched uranium oxycarbide (UCO). Kernel diameters are approximately 350 µm with a U-235 enrichment of approximately 19.7%. Compact 6-3-2more » has been irradiated to 11.3% FIMA compact average burn-up with a time average, volume average temperature of 1070.2°C and with a compact average fast fluence of 2.38E21 n/cm« less

Electron Microscopic Examination of Irradiated TRISO Coated Particles of Compact 6-3-2 of AGR-1 Experiment

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

Van Rooyen, Isabella Johanna; Demkowicz, Paul Andrew; Riesterer, Jessica Lori

2012-12-01

The electron microscopic examination of selected irradiated TRISO coated particles of the AGR-1 experiment of fuel compact 6-3-2 are presented in this report. Compact 6-3-2 refers to the compact in Capsule 6 at level 3 of Stack 2. The fuel used in capsule 6 compacts, are called the “baseline” fuel as it is fabricated with refined coating process conditions used to fabricate historic German fuel, because of its excellent irradiation performance with UO 2 kernels. The AGR-1 fuel is however made of low-enriched uranium oxycarbide (UCO). Kernel diameters are approximately 350 µm with a U-235 enrichment of approximately 19.7%. Compactmore » 6-3-2 has been irradiated to 11.3% FIMA compact average burn-up with a time average, volume average temperature of 1070.2°C and with a compact average fast fluence of 2.38E21 n/cm« less

Nuclear Fuel Reprocessing

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

Harold F. McFarlane; Terry Todd

2013-11-01

Reprocessing is essential to closing nuclear fuel cycle. Natural uranium contains only 0.7 percent 235U, the fissile (see glossary for technical terms) isotope that produces most of the fission energy in a nuclear power plant. Prior to being used in commercial nuclear fuel, uranium is typically enriched to 3–5% in 235U. If the enrichment process discards depleted uranium at 0.2 percent 235U, it takes more than seven tonnes of uranium feed to produce one tonne of 4%-enriched uranium. Nuclear fuel discharged at the end of its economic lifetime contains less one percent 235U, but still more than the natural ore.more » Less than one percent of the uranium that enters the fuel cycle is actually used in a single pass through the reactor. The other naturally occurring isotope, 238U, directly contributes in a minor way to power generation. However, its main role is to transmute into plutoniumby neutron capture and subsequent radioactive decay of unstable uraniumand neptuniumisotopes. 239Pu and 241Pu are fissile isotopes that produce more than 40% of the fission energy in commercially deployed reactors. It is recovery of the plutonium (and to a lesser extent the uranium) for use in recycled nuclear fuel that has been the primary focus of commercial reprocessing. Uraniumtargets irradiated in special purpose reactors are also reprocessed to obtain the fission product 99Mo, the parent isotope of technetium, which is widely used inmedical procedures. Among the fission products, recovery of such expensive metals as platinum and rhodium is technically achievable, but not economically viable in current market and regulatory conditions. During the past 60 years, many different techniques for reprocessing used nuclear fuel have been proposed and tested in the laboratory. However, commercial reprocessing has been implemented along a single line of aqueous solvent extraction technology called plutonium uranium reduction extraction process (PUREX). Similarly, hundreds of types of reactor fuels have been irradiated for different purposes, but the vast majority of commercial fuel is uranium oxide clad in zirconium alloy tubing. As a result, commercial reprocessing plants have relatively narrow technical requirements for used nuclear that is accepted for processing.« less

Determination of irradiated reactor uranium in soil samples in Belarus using 236U as irradiated uranium tracer.

PubMed

Mironov, Vladislav P; Matusevich, Janna L; Kudrjashov, Vladimir P; Boulyga, Sergei F; Becker, J Sabine

2002-12-01

This work presents experimental results on the distribution of irradiated reactor uranium from fallout after the accident at Chernobyl Nuclear Power Plant (NPP) in comparison to natural uranium distribution in different soil types. Oxidation processes and vertical migration of irradiated uranium in soils typical of the 30 km relocation area around Chernobyl NPP were studied using 236U as the tracer for irradiated reactor uranium and inductively coupled plasma mass spectrometry as the analytical method for uranium isotope ratio measurements. Measurements of natural uranium yielded significant variations of its concentration in upper soil layers from 2 x 10(-7) g g(-1) to 3.4 x 10(-6) g g(-1). Concentrations of irradiated uranium in the upper 0-10 cm soil layers at the investigated sampling sites varied from 5 x 10(-12) g g(-1) to 2 x 10(-6) g g(-1) depending on the distance from Chernobyl NPP. In the majority of investigated soil profiles 78% to 97% of irradiated "Chernobyl" uranium is still contained in the upper 0-10 cm soil layers. The physical and chemical characteristics of the soil do not have any significant influence on processes of fuel particle destruction. Results obtained using carbonate leaching of 236U confirmed that more than 60% of irradiated "Chernobyl" uranium is still in a tetravalent form, ie. it is included in the fuel matrix (non-oxidized fuel UO2). The average value of the destruction rate of fuel particles determined for the Western radioactive trace (k = 0.030 +/- 0.005 yr(-1)) and for the Northern radioactive trace (k = 0.035 + 0.009 yr(-1)) coincide within experimental errors. Use of leaching of fission products in comparison to leaching of uranium for study of the destruction rate of fuel particles yielded poor coincidence due to the fact that use of fission products does not take into account differences in the chemical properties of fission products and fuel matrix (uranium).

Data summary report for fission product release Test VI-7

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

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

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

The role of dispersed particles in strengthening and fracture mechanisms in a Mo-ZrC alloy processed by mechanical alloying

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

Takida, T.; Mabuchi, M.; Nakamura, M.

2000-03-01

The tensile properties of a ZrC particle-dispersed Mo, which was processed by spark plasma sintering with mechanically alloyed powder, were investigated at room temperature and at elevated temperatures of 1,170 to 1,970 K. The Mo-ZrC alloy showed much higher strength at room temperature than a fully recrystallized pure Mo. The high strength of Mo-ZrC is mainly attributed to a very small grain size (about 3 {micro}m). The main role of the ZrC particle is not to increase strength due to the particle-dislocation interaction, but to limit grain growth during sintering and to attain the very small grain size. The elongationmore » at room temperature of No-ZrC was much lower than that of pure Mo. This is probably related to the higher interstitial contents. However, Mo-ZrC showed a large elongation of 180 pct at 1,970 K and 6.7 x 10{sup {minus}4} s{sup {minus}1}. It was suggested that the ZrC particles stabilized the fine-grained microstructure yet provided no cavitation sites at 1,970 K; as a result, the large elongation was attained.« less

Impurity precipitation in atomized particles evidenced by nano x-ray diffraction computed tomography

NASA Astrophysics Data System (ADS)

Bonnin, Anne; Wright, Jonathan P.; Tucoulou, Rémi; Palancher, Hervé

2014-08-01

Performances and physical properties of high technology materials are influenced or even determined by their initial microstructure and by the behavior of impurity phases. Characterizing these impurities and their relations with the surrounding matrix is therefore of primary importance but it unfortunately often requires a destructive approach, with the risk of misinterpreting the observations. The improvement we have done in high resolution X-ray diffraction computed tomography combined with the use of an X-ray nanoprobe allows non-destructive crystallographic description of materials with microscopic heterogeneous microstructure (with a grain size between 10 nm and 10 μm). In this study, the grain localization in a 2D slice of a 20 μm solidified atomized γU-Mo particle is shown and a minority U(C,O) phase (1 wt. %) with sub-micrometer sized grains was characterized inside. Evidence is presented showing that the onset of U(C,O) grain crystallization can be described by a precipitation mechanism since one single U-Mo grain has direct orientation relationship with more than one surrounding U(C,O) grains.

Fabrication of micro-cell UO2-Mo pellet with enhanced thermal conductivity

NASA Astrophysics Data System (ADS)

Kim, Dong-Joo; Rhee, Young Woo; Kim, Jong Hun; Kim, Keon Sik; Oh, Jang Soo; Yang, Jae Ho; Koo, Yang-Hyun; Song, Kun-Woo

2015-07-01

As one of accident tolerant fuel pellets which should have features of good thermal conductivity and high fission product retention, a micro-cell UO2-Mo pellet has been studied in the aspect of fabrication and thermal property. It was intended to develop the compatible process with conventional UO2 pellet fabrication process. The effects of processing parameters such as the size and density of UO2 granule and the size of Mo powder have been studied to produce sound and dense pellet with completely connected uniform Mo cell-walls. The micro-cell UO2-Mo pellet consists of many Mo micro-cells and UO2 in them. The thermal conductivity of the micro-cell UO2-Mo pellet was measured and compared to those of the UO2 pellet and the UO2-Mo pellet with dispersed form of Mo particles. The thermal conductivity of the micro-cell UO2-Mo pellet was much enhanced and was found to be influenced by the Mo volumetric fraction and pellet integrity. A continuous Mo micro-cell works as a heat conducting channel in the pellet, greatly enhancing the thermal conductivity of the micro cell UO2-Mo pellet.

Activating Aluminum Reactivity with Fluoropolymer Coatings for Improved Energetic Composite Combustion.

PubMed

McCollum, Jena; Pantoya, Michelle L; Iacono, Scott T

2015-08-26

Aluminum (Al) particles are passivated by an aluminum oxide (Al2O3) shell. Energetic blends of nanometer-sized Al particles with liquid perfluorocarbon-based oxidizers such as perfluoropolyethers (PFPE) excite surface exothermic reaction between fluorine and the Al2O3 shell. The surface reaction promotes Al particle reactivity. Many Al-fueled composites use solid oxidizers that induce no Al2O3 surface exothermicity, such as molybdenum trioxide (MoO3) or copper oxide (CuO). This study investigates a perfluorinated polymer additive, PFPE, incorporated to activate Al reactivity in Al-CuO and Al-MoO3. Flame speeds, differential scanning calorimetry (DSC), and quadrupole mass spectrometry (QMS) were performed for varying percentages of PFPE blended with Al/MoO3 or Al/CuO to examine reaction kinetics and combustion performance. X-ray photoelectron spectroscopy (XPS) was performed to identify product species. Results show that the performance of the thermite-PFPE blends is highly dependent on the bond dissociation energy of the metal oxide. Fluorine-Al-based surface reaction with MoO3 produces an increase in reactivity, whereas the blends with CuO show a decline when the PFPE concentration is increased. These results provide new evidence that optimizing Al combustion can be achieved through activating exothermic Al surface reactions.

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

NASA Astrophysics Data System (ADS)

Kleykamp, H.

1997-09-01

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

TEM and XAS investigation of fission gas behaviors in U-Mo alloy fuels through ion beam irradiation

NASA Astrophysics Data System (ADS)

Zang, Hang; Yun, Di; Mo, Kun; Wang, Kunpeng; Mohamed, Walid; Kirk, Marquis A.; Velázquez, Daniel; Seibert, Rachel; Logan, Kevin; Terry, Jeffrey; Baldo, Peter; Yacout, Abdellatif M.; Liu, Wenbo; Zhang, Bo; Gao, Yedong; Du, Yang; Liu, Jing

2017-10-01

In this study, smaller-grained (hundred nano-meter size grain) and larger-grained (micro-meter size grain) U-10Mo specimens have been irradiated (implanted) with 250 keV Xe+ beam and were in situ characterized by TEM. Xe bubbles were not seen in the specimen after an implantation fluence of 2 × 1020 ions/m2 at room temperature. Nucleation of Xe bubbles happened during heating of the specimen to a final temperature of 300 °C. By comparing measured Xe bubble statistics, the nucleation and growth behaviors of Xe bubbles were investigated in smaller-grained and larger-grained U-10Mo specimens. A multi-atom kind of nucleation mechanism has been observed in both specimens. X-ray Absorption spectroscopy showed the edge position in the bubbles to be the same as that of Xe gas. The size of Xe bubbles has been shown to be bigger in larger-grained specimens than in smaller-grained specimens at the same implantation conditions.

The MoEDAL Experiment at the LHC

NASA Astrophysics Data System (ADS)

Pinfold, James L.

2014-04-01

In 2010 the CERN (European Centre for Particle Physics Research) Research Board unanimously approved MoEDAL, the 7th international experiment at the Large Hadron Collider (LHC), which is designed to search for avatars of new physics signified by highly ionizing particles. The MoEDAL detector is like a giant camera ready to reveal "photographic" evidence for new physics and also to actually trap long-lived new particles for further study. The MoEDAL experiment will significantly expand the horizon for discovery at the LHC, in a complementary way. A MoEDAL discovery would have revolutionary implications for our understanding of the microcosm, providing insights into such fundamental questions as: do magnetic monopoles exist, are there extra dimensions or new symmetries of nature; what is the mechanism for the generation of mass; what is the nature of dark matter and how did the big-bang unfurl at the earliest times.

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.

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

Joshi, Vineet V.; Paxton, Dean M.; Lavender, Curt A.

Over the past several years Pacific Northwest National Laboratory (PNNL) has been actively involved in supporting the U.S. Department of Energy National Nuclear Security Administration Office of Material Management and Minimization (formerly Global Threat Reduction Initiative). The U.S. High- Power Research Reactor (USHPRR) project is developing alternatives to existing highly enriched uranium alloy fuel to reduce the proliferation threat. One option for a high-density metal fuel is uranium alloyed with 10 wt% molybdenum (U-10Mo). Forming the U-10Mo fuel plates/foils via rolling is an effective technique and is actively being pursued as part of the baseline manufacturing process. The processing ofmore » these fuel plates requires systematic investigation/understanding of the pre- and post-rolling microstructure, end-state mechanical properties, residual stresses, and defects, their effect on the mill during processing, and eventually, their in-reactor performance. In the work documented herein, studies were conducted to determine the effect of cold and hot rolling the as-cast and homogenized U-10Mo on its microstructure and hardness. The samples were homogenized at 900°C for 48 h, then later annealed for several durations and temperatures to investigate the effect on the material’s microstructure and hardness. The rolling of the as-cast plate, both hot and cold, was observed to form a molybdenum-rich and -lean banded structure. The cold rolling was ineffective, and in some cases exacerbated the as-cast defects. The grains elongated along the rolling direction and formed a pancake shape, while the carbides fractured perpendicularly to the rolling direction and left porosity between fractured particles of UC. The subsequent annealing of these samples at sub-eutectoid temperatures led to rapid precipitation of the ' lamellar phase, mainly in the molybdenum-lean regions. Annealing the samples above the eutectoid temperature did not refine the grain size or the banded microstructure. However, annealing the samples led to quick recovery in hardness as evidenced by a drop in Vickers hardness of 20%. Hot rolling was performed at 650 and 800°C. The hot-rolling mill loads (load separation force) were approximately 40 to 50% less than the cold-rolling for the same reduction and thickness. It was observed that hot rolling the samples with 50% or more reduction in thickness were responsible for dynamic recrystallization in the hot-rolled samples and led to grain refinement. Unlike the cold-rolled samples, the hot-rolled samples did not fracture the carbides and appeared to heal the casting defects. The recovery phenomenon was similar to the cold-rolled samples above the eutectoid temperatures, but owing to the refined grain size, the precipitation of the lamellar phase was far more rapid in these samples and the hardness increased more rapidly than in the cold rolled sample when heated below the eutectoid temperature. The data generated from these rolling efforts has been used to make the process modeling efforts more robust and applicable to all USHPRR partner rolling mills. The flow stress for cold rolling the samples was determined to be between 170-190 ksi, with frictional forces between 0.2 and 0.4 for the PNNL mill. The measured roll separation forces and those simulated using finite element methods for hot and cold rolling for the PNNL rolling mill were in good agreement.« less

A study of ambient fine particles at Tianjin International Airport, China.

PubMed

Ren, Jianlin; Liu, Junjie; Li, Fei; Cao, Xiaodong; Ren, Shengxiong; Xu, Bin; Zhu, Yifang

2016-06-15

The total count number concentration of particles from 10 to 1000nm, particle size distribution, and PM2.5 (aerodynamic diameter≤2.5μm) mass concentration were measured on a parking apron next to the runway at Tianjin International Airport in China. The data were collected 250, 270, 300, 350, and 400m from the runway. Wind direction and wind speed played important roles in determining the characteristics of the atmospheric particles. An inverted U-shaped relationship was observed between the measured particle number concentration and wind speed, with an average peak concentration of 2.2×10(5)particles/cm(3) at wind speeds of approximately 4-5m/s. The atmospheric particle number concentration was affected mainly by aircraft takeoffs and landings, and the PM2.5 mass concentration was affected mainly by the relative humidity (RH) of the atmosphere. Ultrafine particles (UFPs, diameter<100nm), with the highest number concentration at a particle size of approximately 16nm, dominated the measured particle size distributions. The calculated particle emission index values for aircraft takeoff and landing were nearly the same, with mean values of 7.5×10(15)particles/(kg fuel) and 7.6×10(15)particles/(kg fuel), respectively. The particle emission rate for one aircraft during takeoff is two orders of magnitude higher than for all gasoline-powered passenger vehicles in Tianjin combined. The particle number concentrations remained much higher than the background concentrations even beyond 400m from the runway. Copyright © 2016 Elsevier B.V. All rights reserved.

CORROSION STUDIES FOR A FUSED SALT-LIQUID METAL EXTRACTION PROCESS FOR THE LIQUID METAL FUEL REACTOR

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

Susskind, H.; Hill, F.B.; Green, L.

1960-06-30

Corrosion screening tests were carried out on potential materials of construction for use in a fused salt-liquid metal extraction process plant. The corrodents of interest were NaCl--KCl-- MgCl/sub 2/ eutectic, LiCl--KCl eutectic, Bi-- U fuel, and BiCl/sub 3/, either separately or in various combinations. Screening tests to determine the resistance of a wide range of commercial alloys to the corrodents were performed in static and tilting-furnace capsules. Some ceramic materials were tested in static capsules. Largerscale tests of metallic materials were conducted in thermal convection loops and in a forced circulation loop. Some of the tests were conducted isothermally atmore » 500 deg C, and others were performed under 40 to 50 deg C temperature differences at roughly the same teinperature level. On the basis of metallographic examination of exposed test tabs and chemical analyses of corrodents, it was found that the binary and ternary eutectics by themselves produced little attack on any of the materials tested. A wide variety of materials including 1020 mild steel, 2 1/4 Cr--1 Mo alloy steel, types 304 (ELC), 310, 316, 347, 430, and 446 stainless steel, 16-1 Croloy, Inconel, Hastelloy C, Inor-8, Mo, and Ta is, therefore, available for further study. Corrosion by the ternary salt-fuel system was characteristic of that produced by the fuel alone. Alloys such as 1020 mild steel, and 1 1/4 Cr--1/ 2 Mo, and 2 1/4 Cr--1 Mo alloy steel, which are resistant to fuel, would be likely choices at present for container materials. BiCl/sub 3/ produced extensive attack on ternary salt-fuel containers when the fuel contained insufficient concentrations of oxidizable solutes. Au and Al/sub 2/O/sub 3/ were the only materials not attacked by BiCl/sub 3/ in ternary salt alone. (auth)« less

Effects of heat treatment on U–Mo fuel foils with a zirconium diffusion barrier

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

Jue, Jan-Fong; Trowbridge, Tammy L.; Breckenridge, Cynthia R.

A monolith fuel design based on U–Mo alloy has been selected as the fuel type for conversion of the United States’ high performance research reactors (HPRRs) from highly enriched uranium (HEU) to low-enriched uranium (LEU). In this fuel design, a thin layer of zirconium is used to eliminate the direct interaction between the U–Mo fuel meat and the aluminum-alloy cladding during irradiation. The co-rolling process used to bond the Zr barrier layer to the U–Mo foil during fabrication alters the microstructure of both the U–10Mo fuel meat and the U–Mo/Zr interface. This work studied the effects of post-rolling annealing treatmentmore » on the microstructure of the co-rolled U–Mo fuel meat and the U–Mo/Zr interaction layer. Microscopic characterization shows that the grain size of U–Mo fuel meat increases with the annealing temperature, as expected. The grain sizes were ~9, ~13, and ~20 μm for annealing temperature of 650, 750, and 850 °C, respectively. No abnormal grain growth was observed. The U–Mo/Zr interaction-layer thickness increased with the annealing temperature with an Arrhenius constant for growth of 184 kJ/mole, consistent with a previous diffusion-couple study. The interaction layer thickness was 3.2 ± 0.5 μm, 11.1 ± 2.1 μm, 27.1 ± 0.9 μm for annealing temperature of 650, 750, to 850 °C, respectively. The homogeneity of Mo improves with post rolling annealing temperature and with U–Mo coupon homogenization. The phases in the Zr/U–Mo interaction layer produced by co-rolling, however, differ from those reported in the previous diffusion couple studies.« less

In situ measurement of low-Z material coating thickness on high Z substrate for tokamaks.

PubMed

Mueller, D; Roquemore, A L; Jaworski, M; Skinner, C H; Miller, J; Creely, A; Raman, P; Ruzic, D

2014-11-01

Rutherford backscattering of energetic particles can be used to determine the thickness of a coating of a low-Z material over a heavier substrate. Simulations indicate that 5 MeV alpha particles from an (241)Am source can be used to measure the thickness of a Li coating on Mo tiles between 0.5 and 15 μm thick. Using a 0.1 mCi source, a thickness measurement can be accomplished in 2 h of counting. This technique could be used to measure any thin, low-Z material coating (up to 1 mg/cm(2) thick) on a high-Z substrate, such as Be on W, B on Mo, or Li on Mo. By inserting a source and detector on a moveable probe, this technique could be used to provide an in situ measurement of the thickness of Li coating on NSTX-U Mo tiles. A test stand with an alpha source and an annular solid-state detector was used to investigate the measurable range of low-Z material thicknesses on Mo tiles.

In situ measurement of low-Z material coating thickness on high Z substrate for tokamaks

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

Mueller, D., E-mail: dmueller@pppl.gov; Roquemore, A. L.; Jaworski, M.

Rutherford backscattering of energetic particles can be used to determine the thickness of a coating of a low-Z material over a heavier substrate. Simulations indicate that 5 MeV alpha particles from an {sup 241}Am source can be used to measure the thickness of a Li coating on Mo tiles between 0.5 and 15 μm thick. Using a 0.1 mCi source, a thickness measurement can be accomplished in 2 h of counting. This technique could be used to measure any thin, low-Z material coating (up to 1 mg/cm{sup 2} thick) on a high-Z substrate, such as Be on W, B on Mo, or Limore » on Mo. By inserting a source and detector on a moveable probe, this technique could be used to provide an in situ measurement of the thickness of Li coating on NSTX-U Mo tiles. A test stand with an alpha source and an annular solid-state detector was used to investigate the measurable range of low-Z material thicknesses on Mo tiles.« less

Uranium nitride as LWR TRISO fuel: Thermodynamic modeling of U-C-N

NASA Astrophysics Data System (ADS)

Besmann, Theodore M.; Shin, Dongwon; Lindemer, Terrence B.

2012-08-01

TRISO coated particle fuel is envisioned as a next generation replacement for current urania pellet fuel in LWR applications. To obtain adequate fissile loading the kernel of the TRISO particle will likely need to be UN instead of UO2. In support of the necessary development effort for this new fuel system, an assessment of phase regions of interest in the U-C-N system was undertaken as the fuel will be prepared by the carbothermic reduction of the oxide followed by nitriding, will be in equilibrium with carbon within the TRISO particle, and will react with minor actinides and fission products. The phase equilibria and thermochemistry of the U-C-N system is reviewed, including nitrogen pressure measurements above various phase fields. Measurements were used to confirm an ideal solution model of UN and UC adequately represents the UC1-xNx phase. Agreement with the data was significantly improved by effectively adjusting the Gibbs free energy of UN by +12 kJ/mol. This also required adjustment of the value for the sesquinitride by +17 kJ/mol to obtain agreement with phase equilibria. The resultant model together with reported values for other phases in the system was used to generate isothermal sections of the U-C-N phase diagram. Nitrogen partial pressures were also computed for regions of interest.

Estimating the future agriculture freight transportation network needs due to climate change using remote sensing and regional climate models.

DOT National Transportation Integrated Search

2016-12-01

A reoccurring challenge with increasing fuel prices is optimization of multi- and inter-modal freight transport to move products most efficiently. Projections for the future of agriculture in the United States (U.S.) combined with regional climate mo...

Investigation of the effects of radiolytic-gas bubbles on the long-term operation of solution reactors for medical-isotope production

NASA Astrophysics Data System (ADS)

Souto Mantecon, Francisco Javier

One of the most common and important medical radioisotopes is 99Mo, which is currently produced using the target irradiation technology in heterogeneous nuclear reactors. The medical isotope 99Mo can also be produced from uranium fission using aqueous homogeneous solution reactors. In solution reactors, 99Mo is generated directly in the fuel solution, resulting in potential advantages when compared with the target irradiation process in heterogeneous reactors, such as lower reactor power, less waste heat, and reduction by a factor of about 100 in the generation of spent fuel. The commercial production of medical isotopes in solution reactors requires steady-state operation at about 200 kW. At this power regime, the formation of radiolytic-gas bubbles creates a void volume in the fuel solution that introduces a negative coefficient of reactivity, resulting in power reduction and instabilities that may impede reactor operation for medical-isotope production. A model has been developed considering that reactivity effects are due to the increase in the fuel-solution temperature and the formation of radiolytic-gas bubbles. The model has been validated against experimental results from the Los Alamos National Laboratory uranyl fluoride Solution High-Energy Burst Assembly (SHEBA), and the SILENE uranyl nitrate solution reactor, commissioned at the Commissariat a l'Energie Atomique, in Valduc, France. The model shows the feasibility of solution reactors for the commercial production of medical isotopes and reveals some of the important parameters to consider in their design, including the fuel-solution type, 235U enrichment, uranium concentration, reactor vessel geometry, and neutron reflectors surrounding the reactor vessel. The work presented herein indicates that steady-state operation at 200 kW can be achieved with a solution reactor consisting of 120 L of uranyl nitrate solution enriched up to 20% with 235U and a uranium concentration of 145 kg/m3 in a graphite-reflected cylindrical geometry.

Manufacturing Experience for Oxide Dispersion Strengthened Alloys

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

Bennett, Wendy D.; Doherty, Ann L.; Henager, Charles H.

2016-09-22

This report documents the results of the development and the manufacturing experience gained at the Pacific Northwest National Laboratories (PNNL) while working with the oxide dispersion strengthened (ODS) materials MA 956, 14YWT, and 9YWT. The Fuel Cycle Research and Development program of the Office of Nuclear Energy has implemented a program to develop a Uranium-Molybdenum metal fuel for light water reactors. ODS materials have the potential to provide improved performance for the U-Mo concept.

Post-irradiation-examination of irradiated fuel outside the hot cell

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

Dawn E. Janney; Adam B. Robinson; Thomas P. O'Holleran

Because of their high radioactivity, irradiated fuels are commonly examined in a hot cell. However, the Idaho National Laboratory (INL) has recently investigated irradiated U-Mo-Al metallic fuel from the Reduced Enrichment for Research and Test Reactors (RERTR) project using a conventional unshielded scanning electron microscope outside a hot cell. This examination was possible because of a two-step sample-preparation approach in which a small volume of fuel was isolated in a hot cell and shielding was introduced during later stages of sample preparation. The resulting sample contained numerous sample-preparation artifacts but allowed analysis of microstructures from selected areas.

A Metal Stable Isotope Approach to Understanding Uranium Mobility Across Roll Front Redox Boundaries

NASA Astrophysics Data System (ADS)

Brown, S. T.; Basu, A.; Christensen, J. N.; DePaolo, D. J.; Heikoop, J. M.; Reimus, P. W.; Maher, K.; Weaver, K. L.

2015-12-01

Sedimentary roll-front uranium (U) ore deposits are the principal source of U for nuclear fuel in the USA and an important part of the current all-of-the-above energy strategy. Mining of roll-front U ore in the USA is primarily by in situ alkaline oxidative dissolution of U minerals. There are significant environmental benefits to in situ mining including no mine tailings or radioactive dust, however, the long-term immobilization of U in the aquifer after the completion of mining remains uncertain. We have utilized the metal stable isotopes U, Se and Mo in groundwater from roll-front mines in Texas and Wyoming to quantify the aquifer redox conditions and predict the onset of U reduction after post mining aquifer restoration. Supporting information from the geochemistry of groundwater and aquifer sediments are used to understand the transport of U prior to and after in situ mining. Groundwater was collected across 4 mining units at the Rosita mine in the Texas coastal plain and 2 mining units at the Smith Ranch mine in the Powder River Basin, Wyoming. In general, the sampled waters are moderately reducing and ore zone wells contain the highest aqueous U concentrations. The lowest U concentrations occur in monitoring wells downgradient of the ore zone. 238U/235U is lowest in downgradient wells and is correlated with aqueous U concentrations. Rayleigh distillation models of the 238U/235U are consistent with U isotope fractionation factors of 1.0004-1.001, similar to lab-based studies. Based on these results we conclude that redox reactions continue to affect U distribution in the ore zone and downgradient regions. We also measured aqueous selenium isotope (δ82Se) and molybdenum isotope (δ98Mo) compositions in the Rosita groundwater. Se(VI) primarily occurs in the upgradient wells and is absent in most ore zone and downgradient wells. Rayleigh distillation models suggest reduction of Se(VI) along the groundwater flow path and when superimposed on the U isotope data Se reduction is favored over U reduction. The δ98Mo of Rosita groundwater is significantly elevated compared to the U ore and is negatively correlated with the groundwater Eh, which suggests localized strong reducing conditions capable of Mo reduction. Ongoing work will determine the Mo isotope systematics of U ores and groundwater from roll-front deposits.

The structure of particle cloud premixed flames

NASA Technical Reports Server (NTRS)

Seshadri, K.; Berlad, A. L.

1992-01-01

The structure of premixed flames propagating in combustible systems containing uniformly distributed volatile fuel particles in an oxidizing gas mixture is analyzed. This analysis is motivated by experiments conducted at NASA Lewis Research Center on the structure of flames propagating in combustible mixtures of lycopodium particles and air. Several interesting modes of flame propagation were observed in these experiments depending on the number density and the initial size of the fuel particle. The experimental results show that steady flame propagation occurs even if the initial equivalence ratio of the combustible mixture based on the gaseous fuel available in the particles, phi sub u, is substantially larger than unity. A model is developed to explain these experimental observations. In the model, it is presumed that the fuel particles vaporize first to yield a gaseous fuel of known chemical composition which then reacts with oxygen in a one-step overall process. The activation energy of the chemical reaction is presumed to be large. The activation energy characterizing the kinetics of vaporization is also presumed to be large. The equations governing the structure of the flame were integrated numerically. It is shown that the interplay of vaporization kinetics and oxidation process can result in steady flame propagation in combustible mixtures where the value of phi sub u is substantially larger than unity. This prediction is in agreement with experimental observations.

Optimisation of composite metallic fuel for minor actinide transmutation in an accelerator-driven system

NASA Astrophysics Data System (ADS)

Uyttenhove, W.; Sobolev, V.; Maschek, W.

2011-09-01

A potential option for neutralization of minor actinides (MA) accumulated in spent nuclear fuel of light water reactors (LWRs) is their transmutation in dedicated accelerator-driven systems (ADS). A promising fuel candidate dedicated to MA transmutation is a CERMET composite with Mo metal matrix and (Pu, Np, Am, Cm)O 2-x fuel particles. Results of optimisation studies of the CERMET fuel targeting to increasing the MA transmutation efficiency of the EFIT (European Facility for Industrial Transmutation) core are presented. In the adopted strategy of MA burning the plutonium (Pu) balance of the core is minimized, allowing a reduction in the reactivity swing and the peak power form-factor deviation and an extension of the cycle duration. The MA/Pu ratio is used as a variable for the fuel optimisation studies. The efficiency of MA transmutation is close to the foreseen theoretical value of 42 kg TW -1 h -1 when level of Pu in the actinide mixture is about 40 wt.%. The obtained results are compared with the reference case of the EFIT core loaded with the composite CERCER fuel, where fuel particles are incorporated in a ceramic magnesia matrix. The results of this study offer additional information for the EFIT fuel selection.

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

Rabin, S.A.; Martin, M.M.; Lotts, A.L.

The fabricability of dispersion fuels using UO/sub 2/ or UC as the dispersoid and uranium combined with 10 to 15 wt% Mo as the matrix was investigated. Cores containing l7.8 wt% UO/sub 2/ dispersed in U-- 15 wt.% Mo were successfully fabricated to about 80% of theoretical density by cold pressing at 50 tsi, sintering at 1100 deg C, and cold coining at 50 tsi. Comparable results were obtained with UC as the dispersoid. Core fabrication results varied greatly with the type of matrix powder used. Occluded gases, pour density, and surface cleanliness bore important relations to the fabrication behaviormore » of powders. Suitable pressing and sintering results were obtained with prealloyed, calcium-reduced U--Mo powder and with molybdenum and calcium-reduced uranium as elemental powders. Shotted prealloyed powders were difficult to press and sinter, as were elemental and prealloyed powders prepared by hydriding. The cores containing UO/sub 2/ were picture-frame, hot-roll-clad as miniature plates. Molybdenum, Fansteel 82, and Zr--3 wt% Al were investigated as cladding materials. While each bonded well to itself, only the molybdenum-clad core, rolled at 1150 deg C to 10/1 reduction, resulted in dispersions free of ruptures and UO/sub 2/ fragmentation and in strong bonding to the core, evaluated by metallography, mechanical peel, and thermal shock tests. The matrix phase was homogeneous, but the UO/sub 2/ dispersoid showed stringering characteristic of cores worked by hot rolling. Core densities as high as 99% of theoretical were obtained. (auth)« less

Optimizing catalysis conditions to decrease aromatic hydrocarbons and increase alkanes for improving jet biofuel quality.

PubMed

Cheng, Jun; Li, Tao; Huang, Rui; Zhou, Junhu; Cen, Kefa

2014-04-01

To produce quality jet biofuel with high amount of alkanes and low amount of aromatic hydrocarbons, two zeolites of HY and HZSM-5 supporting Ni and Mo were used as catalysts to convert soybean oil into jet fuel. Zeolite HY exhibited higher jet range alkane selectivity (40.3%) and lower jet range aromatic hydrocarbon selectivity (23.8%) than zeolite HZSM-5 (13.8% and 58.9%). When reaction temperature increased from 330 to 390°C, yield of jet fuel over Ni-Mo/HY catalyst at 4 MPa hydrogen pressure increased from 0% to 49.1% due to the shift of reaction pathway from oligomerization to cracking reaction. Further increase of reaction temperature from 390 to 410°C resulted in increased yield of jet range aromatic hydrocarbons from 18.7% to 30%, which decreased jet fuel quality. A high yield of jet fuel (48.2%) was obtained at 1 MPa low hydrogen pressure over Ni (8 wt.%)-Mo (12 wt.%)/HY catalyst. Copyright © 2014 Elsevier Ltd. All rights reserved.

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

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

Rest, J.; Hofman, G.L.

1997-06-01

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

Neutronics and Thermal Hydraulics Study for Using a Low-Enriched Uranium Core in the Advanced Test Reactor -- 2008 Final Report

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

G. S. Chang; M. A. Lillo; R. G. Ambrosek

2008-06-01

The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuelmore » cycle burnup comparison analysis. Using the current HEU U 235 enrichment of 93.0 % as a baseline, an analysis was performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff versus effective full power days (EFPDs) between the HEU and the LEU cores. The MCNP ATR 1/8th core model was used to optimize the U 235 loading in the LEU core, such that the differences in K-eff and heat flux profiles between the HEU and LEU cores were minimized. The depletion methodology MCWO was used to calculate K-eff versus EFPDs in this paper. The MCWO-calculated results for the LEU demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar to the ATR reference HEU case study. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm (20 mil). In this work, the proposed LEU (U-10Mo) core conversion case with nominal fuel meat thickness of 0.330 mm (13 mil) and U-235 enrichment of 19.7 wt% is used to optimize the radial heat flux profile by varying the fuel meat thickness from 0.191 mm (7.0 mil) to 0.330 mm (13.0 mil) at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19). A 0.8g of Boron-10, a burnable absorber, was added in the inner and outer plates to reduce the initial excess reactivity, and the peak to average ratio of the inner/outer heat flux more effectively. Because the B-10 (n,a) reaction will produce Helium-4 (He-4), which might degrade the LEU foil type fuel performance, an alternative absorber option is proposed. The proposed LEU case study will have 6.918 g of Cadmium (Cd) mixed with the LEU at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19) as a burnable absorber to achieve peak to average ratios similar to those for the ATR reference HEU case study.« less

Non-destructive Quantitative Phase Analysis and Microstructural Characterization of Zirconium Coated U-10Mo Fuel Foils via Neutron Diffraction

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

Cummins, Dustin Ray; Vogel, Sven C.; Hollis, Kendall Jon

2016-10-18

This report uses neutron diffraction to investigate the crystal phase composition of uranium-molybdenum alloy foils (U-10Mo) for the CONVERT MP-1 Reactor Conversion Project, and determines the effect on alpha-uranium contamination following the deposition of a Zr metal diffusion layer by various methods: plasma spray deposition of Zr powders at LANL and hot co-rolling with Zr foils at BWXT. In summary, there is minimal decomposition of the gamma phase U-10Mo foil to alpha phase contamination following both plasma spraying and hot co-rolling. The average unit cell volume, i.e. lattice spacing, of the Zr layer can be mathematically extracted from the diffractionmore » data; co-rolled Zr matches well with literature values of bulk Zr, while plasma sprayed Zr shows a slight increase in the lattice spacing, indicative of interstitial oxygen in the lattice. Neutron diffraction is a beneficial alternative to conventional methods of phase composition, i.e. x ray diffraction (XRD) and destructive metallography. XRD has minimal penetration depth in high atomic number materials, particularly uranium, and can only probe the first few microns of the fuel plate; neutrons pass completely through the foil, allowing for bulk analysis of the foil composition and no issues with addition of cladding layers, as in the final, aluminum-clad reactor fuel plates. Destructive metallography requires skilled technicians, cutting of the foil into small sections, hazardous etching conditions, long polishing and microscopy times, etc.; the neutron diffraction system has an automated sample loader and can fit larger foils, so there is minimal analysis preparation; the total spectrum acquisition time is ~ 1 hour per sample. The neutron diffraction results are limited by spectra refinement/calculation times and the availability of the neutron beam source. In the case of LANSCE at Los Alamos, the beam operates ~50% of the year. Following the lessons learned from these preliminary results, optimizations to the process and analysis can be made, and neutron diffraction can become a viable and efficient technique for gamma/alpha phase composition determination for nuclear fuels.« less

Dissolved molybdenum and uranium in the Three Rivers of eastern Tibet

NASA Astrophysics Data System (ADS)

Noh, H.; Huh, Y.

2006-12-01

Three large rivers - the Chang Jiang (Yangtze), Mekong (Lancang Jiang) and Salween (Nu Jiang) - originate in eastern Tibet and run in close parallel over 300 km near the eastern Himalayan syntaxis. They flow across suture zones and faults generated by the collision of India and Eurasia. Sixty-five water samples were collected in summer of 1999 to 2004 and nine in winter of 2002 to 2003. The complex geologic makeup of the Three Rivers region (TRR) results in widely varying major and trace element compositions of the dissolved load. Two redox-sensitive elements, molybdenum (Mo) and uranium (U) were analyzed by ICP-MS, as potential proxies for weathering of sedimentary organic carbon and resultant generation of atmospheric carbon dioxide. Additionally, Mo constitutes an essential co-enzyme for biology. Mo concentration ranges from 0.76 to 21.3 nmol/kg (average: 6.24 nmol/kg, average of global rivers: ~5 nmol/kg (Martin and Meybeck, 1979)), and U concentration varies from 2.86 to 10.7 nmol/kg (average: 3.12 nmol/kg, average of global rivers: ~1 nmol/kg (Palmer and Edmond, 1993)). The highest values of Mo and U are observed in the headwater tributary sample of the Chang Jiang, where evaporite dissolution is dominant. Statistical analyses show that Mo is closely correlated with U (r = 0.713, p < 0.01) indicating similar source of Mo and U to river waters in the TRR. Inverse correlation with Si/total anions ratio suggests that their sources are non-silicate minerals. The correlation with sulfate supports the use Mo and U as proxies for weathering of reduced organic-rich sediments (Mo and SO4: r = 0.383, p < 0.01; U and SO4: r = 0.508, p < 0.01). Among the parameters tested (basin area, elevation, relief, slope, T, precipitation, potential-evapotranspiration, normalized difference vegetation index (NDVI), population density), best positive correlation (r>0.5) is shown between U and basin elevation, and negative correlation is shown between U and temperature, precipitation and potential- evapotranspiration. Martin J.-M. and Meybeck M. (1979) Elemental mass-balance of material carried by major world rivers. Mar. Chem. 7, 173-206. Palmer M. R. and Edmond J. M. (1993) Uranium in river water. Geochim. Cosmochim. Acta 57, 4947-4955.

Uranium extraction from TRISO-coated fuel particles using supercritical CO2 containing tri-n-butyl phosphate.

PubMed

Zhu, Liyang; Duan, Wuhua; Xu, Jingming; Zhu, Yongjun

2012-11-30

High-temperature gas-cooled reactors (HTGRs) are advanced nuclear systems that will receive heavy use in the future. It is important to develop spent nuclear fuel reprocessing technologies for HTGR. A new method for recovering uranium from tristructural-isotropic (TRISO-) coated fuel particles with supercritical CO(2) containing tri-n-butyl phosphate (TBP) as a complexing agent was investigated. TRISO-coated fuel particles from HTGR fuel elements were first crushed to expose UO(2) pellet fuel kernels. The crushed TRISO-coated fuel particles were then treated under O(2) stream at 750°C, resulting in a mixture of U(3)O(8) powder and SiC shells. The conversion of U(3)O(8) into solid uranyl nitrate by its reaction with liquid N(2)O(4) in the presence of a small amount of water was carried out. Complete conversion was achieved after 60 min of reaction at 80°C, whereas the SiC shells were not converted by N(2)O(4). Uranyl nitrate in the converted mixture was extracted with supercritical CO(2) containing TBP. The cumulative extraction efficiency was above 98% after 20 min of online extraction at 50°C and 25 MPa, whereas the SiC shells were not extracted by TBP. The results suggest an attractive strategy for reprocessing spent nuclear fuel from HTGR to minimize the generation of secondary radioactive waste. Copyright © 2012 Elsevier B.V. All rights reserved.

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

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

Harms, Gary A.

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

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

Mitchell K Meyer

Blister–threshold testing of fuel plates is a standard method through which the safety margin for operation of plate-type in research and test reactors is assessed. The blister-threshold temperature is indicative of the ability of fuel to operate at high temperatures for short periods of time (transient conditions) without failure. This method of testing was applied to the newly developed U-Mo monolithic fuel system. Blister annealing studies on the U-Mo monolithic fuel plates began in 2007, with the Reduced Enrichment for Research and Test Reactors (RERTR)-6 experiment, and they have continued as the U-Mo fuel system has evolved through the researchmore » and development process. Blister anneal threshold temperatures from early irradiation experiments (RERTR-6 through RERTR-10) ranged from 400 to 500°C. These temperatures were projected to be acceptable for NRC-licensed research reactors and the high-power Advanced Test Reactor (ATR) and the High Flux Isotope Reactor (HFIR) based on current safety-analysis reports (SARs). Initial blister testing results from the RERTR-12 experiment capsules X1 and X2 showed a decrease in the blister-threshold temperatures. Blister threshold temperatures from this experiment ranged from 300 to 400°C. Selected plates from the AFIP-4 experiment, which was fabricated using a process similar to that used to fabricate the RERTR-12 experiment, also underwent blister testing to determine whether results would be similar. The measured blister-threshold temperatures from the AFIP-4 plates fell within the same blister-threshold temperature range measured in the RERTR-12 plates. Investigation of the cause of this decrease in bister threshold temperature is being conducted under the guidance of Idaho National Laboratory PLN-4155, “Analysis of Low Blister Threshold Temperatures in the RERTR-12 and AFIP-4 Experiments,” and is driven by hypotheses. The main focus of the investigation is in the following areas: 1. Fabrication variables 2. Pre-irradiation characterization 3. Irradiation conditions 4. Post-irradiation examination 5. Additional blister testing 6. Mechanical modeling This report documents the preliminary results of this investigation. Several hypotheses can be dismissed as a result of this investigation. Two primary categories of causes remain. The most prominent theory, supported by the data, is that low blister-threshold temperature is the result of mechanical energy imparted on the samples during the fabrication process (hot and cold rolling) without adequate post processing (annealing). The mechanisms are not clearly understood and require further investigation, but can be divided into two categories: • Residual Stress • Undesirable interaction boundary and/or U-Mo microstructure change A secondary theory that cannot be dismissed with the information that is currently available is that a change in the test conditions has resulted in a statistically significant downward shift of measured blister temperature. This report outlines the results of the forensic investigations conducted to date. The data and conclusions presented in this report are preliminary. Definitive cause and effect relationships will be established by future experimental programs.« less

First scintillating bolometer tests of a CLYMENE R&D on Li2MoO4 scintillators towards a large-scale double-beta decay experiment

NASA Astrophysics Data System (ADS)

Buşe, G.; Giuliani, A.; de Marcillac, P.; Marnieros, S.; Nones, C.; Novati, V.; Olivieri, E.; Poda, D. V.; Redon, T.; Sand, J.-B.; Veber, P.; Velázquez, M.; Zolotarova, A. S.

2018-05-01

A new R&D on lithium molybdate scintillators has begun within a project CLYMENE (Czochralski growth of Li2MoO4 crYstals for the scintillating boloMeters used in the rare EveNts sEarches). One of the main goals of the CLYMENE is a realization of a Li2MoO4 crystal growth line to be complementary to the one recently developed by LUMINEU in view of a mass production capacity for CUPID, a next-generation tonne-scale bolometric experiment to search for neutrinoless double-beta decay. In the present paper we report the investigation of performance and radiopurity of 158-g and 13.5-g scintillating bolometers based on a first large-mass (230 g) Li2MoO4 crystal scintillator developed within the CLYMENE project. In particular, a good energy resolution (2-7 keV FWHM in the energy range of 0.2-5 MeV), one of the highest light yield (0.97 keV/MeV) amongst Li2MoO4 scintillating bolometers, an efficient alpha particles discrimination (10 σ) and potentially low internal radioactive contamination (below 0.2-0.3 mBq/kg of U/Th, but 1.4 mBq/kg of 210Po) demonstrate prospects of the CLYMENE in the development of high quality and radiopure Li2MoO4 scintillators for CUPID.

Insights on geochemical cycling of U, Re and Mo from seasonal sampling in Boston Harbor, Massachusetts, USA

USGS Publications Warehouse

Morford, J.L.; Martin, W. R.; Kalnejais, Linda H.; Francois, R.; Bothner, Michael H.; Karle, I.-M.

2007-01-01

This study examined the removal of U, Mo, and Re from seawater by sedimentary processes at a shallow-water site with near-saturation bottom water O2 levels (240–380 μmol O2/L), very high organic matter oxidation rates (annually averaged rate is 880 μmol C/cm2/y), and shallow oxygen penetration depths (4 mm or less throughout the year). Under these conditions, U, Mo, and Re were removed rapidly to asymptotic pore water concentrations of 2.2–3.3 nmol/kg (U), 7–13 nmol/kg (Mo), and 11–14 pmol/kg (Re). The depth order in which the three metals were removed, determined by fitting a diffusion-reaction model to measured profiles, was Re < U < Mo. Model fits also suggest that the Mo profiles clearly showed the presence of a near-interface layer in which Mo was added to pore waters by remineralization of a solid phase. The importance of this solid phase source of pore water Mo increased from January to October as the organic matter oxidation rate increased, bottom water O2 decreased, and the O2 penetration depth decreased. Experiments with in situ benthic flux chambers generally showed fluxes of U and Mo into the sediments. However, when the overlying water O2 concentration in the chambers was allowed to drop to very low levels, Mn and Fe were released to the overlying water along with the simultaneous release of Mo and U. These experiments suggest that remineralization of Mn and/or Fe oxides may be a source of Mo and perhaps U to pore waters, and may complicate the accumulation of U and Mo in bioturbated sediments with high organic matter oxidation rates and shallow O2 penetration depths.Benthic chamber experiments including the nonreactive solute tracer, Br−, indicated that sediment irrigation was very important to solute exchange at the study site. The enhancement of sediment–seawater exchange due to irrigation was determined for the nonreactive tracer (Br−), TCO2, NH4+">NH4+, U and Mo. The comparisons between these solutes showed that reactions within and around the burrows were very important for modulating the Mo flux, but less important for U. The effect of these reactions on Mo exchange was highly variable, enhancing Mo (and, to a lesser extent, U) uptake at times of relatively modest irrigation, but inhibiting exchange when irrigation rates were faster. These results reinforce the observation that Mo can be released to and removed from pore waters via sedimentary reactions.The removal rate of U and Mo from seawater by sedimentary reactions was found to agree with the rate of accumulation of authigenic U and Mo in the solid phase. The fluxes of U and Mo determined by in situ benthic flux chamber measurements were the largest that have been measured to date. These results confirm that removal of redox-sensitive metals from continental margin sediments underlying oxic bottom water is important, and suggest that continental margin sediments play a key role in the marine budgets of these metals.

Synthesizing 2D MoS2 Nanofins on carbon nanospheres as catalyst support for Proton Exchange Membrane Fuel Cells.

PubMed

Hu, Yan; Chua, Daniel H C

2016-06-15

Highly dense 2D MoS2 fin-like nanostructures on carbon nanospheres were fabricated and formed the main catalyst support structure in the oxygen reduction reaction (ORR) for polymer electrolyte membrane (PEM) fuel cells. These nanofins were observed growing perpendicular to the carbon nanosphere surface in random orientations and high resolution transmission electron microscope confirmed 2D layers. The PEM fuel cell test showed enhanced electrochemical activity with good stability, generating over 8.5 W.mgPt(-1) as compared to standard carbon black of 7.4 W.mgPt(-1) under normal operating conditions. Electrochemical Impedance Spectroscopy confirmed that the performance improvement is highly due to the excellent water management of the MoS2 lamellar network, which facilitates water retention at low current density and flood prevention at high current density. Reliability test further demonstrated that these nanofins are highly stable in the electrochemical reaction and is an excellent ORR catalyst support.

Synthesizing 2D MoS2 Nanofins on carbon nanospheres as catalyst support for Proton Exchange Membrane Fuel Cells

PubMed Central

Hu, Yan; Chua, Daniel H. C.

2016-01-01

Highly dense 2D MoS2 fin-like nanostructures on carbon nanospheres were fabricated and formed the main catalyst support structure in the oxygen reduction reaction (ORR) for polymer electrolyte membrane (PEM) fuel cells. These nanofins were observed growing perpendicular to the carbon nanosphere surface in random orientations and high resolution transmission electron microscope confirmed 2D layers. The PEM fuel cell test showed enhanced electrochemical activity with good stability, generating over 8.5 W.mgPt−1 as compared to standard carbon black of 7.4 W.mgPt−1 under normal operating conditions. Electrochemical Impedance Spectroscopy confirmed that the performance improvement is highly due to the excellent water management of the MoS2 lamellar network, which facilitates water retention at low current density and flood prevention at high current density. Reliability test further demonstrated that these nanofins are highly stable in the electrochemical reaction and is an excellent ORR catalyst support. PMID:27302135

Results of irradiation of (U0.55Pu0.45)N and (U0.4Pu0.6)N fuels in BOR-60 up to ˜12 at.% burn-up

NASA Astrophysics Data System (ADS)

Rogozkin, B. D.; Stepennova, N. M.; Fedorov, Yu. Ye.; Shishkov, M. G.; Kryukov, F. N.; Kuzmin, S. V.; Nikitin, O. N.; Belyaeva, A. V.; Zabudko, L. M.

2013-09-01

In the article presented are the results of post-irradiation tests of helium bonded fuel pins with mixed mononitride fuel (U0.55Pu0.45)N and (U0.4Pu0.6)N having 85% density irradiated in BOR-60 reactor. Achieved maximum burn-up was, respectively, equal to 9.4 and 12.1 at.% with max linear heat rates 41.9 and 54.5 kW/m. Maximum irradiation dose was 43 dpa. No damage of claddings made of ChS-68 steel (20% cold worked) was observed, and ductility margin existed. Maximum depth of cladding corrosion was within 15 μm. Swelling rates of (U0.4Pu0.6)N and (U0.55Pu0.45)N were, respectively, ˜1.1% and ˜0.68% per 1 at.%. Gas release rate did not exceed 19.3% and 19%. Pattern of porosity distribution in the fuel influenced fuel swelling and gas release rates. Plutonium and uranium are uniformly distributed in the fuel, local minimum values of their content being caused by pores and cracks in the pellets. The observable peaks in content distribution are probably connected with the local formation of isolated phases (e.g. Mo, Pd) while the minimum values refer to fuel pores and cracks. Xenon and cesium tend to migrate from the hot sections of fuel, and therefore their min content is observed in the central section of the fuel pellets. Phase composition of the fuel was determined with X-ray diffractometer. The X-ray patterns of metallographic specimens were obtained by the scanning method (the step was 0.02°, the step exposition was equal to 2 s). From the X-ray diffraction analysis data, it follows that the nitrides of both fuel types have the single-phase structure with an FCC lattice (see Table 6).

Rolling Process Modeling Report: Finite-Element Prediction of Roll Separating Force and Rolling Defects

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

Soulami, Ayoub; Lavender, Curt A.; Paxton, Dean M.

2014-04-23

Pacific Northwest National Laboratory (PNNL) has been investigating manufacturing processes for the uranium-10% molybdenum (U-10Mo) alloy plate-type fuel for the U.S. high-performance research reactors. This work supports the Convert Program of the U.S. Department of Energy’s National Nuclear Security Administration (DOE/NNSA) Global Threat Reduction Initiative. This report documents modeling results of PNNL’s efforts to perform finite-element simulations to predict roll separating forces and rolling defects. Simulations were performed using a finite-element model developed using the commercial code LS-Dyna. Simulations of the hot rolling of U-10Mo coupons encapsulated in low-carbon steel have been conducted following two different schedules. Model predictions ofmore » the roll-separation force and roll-pack thicknesses at different stages of the rolling process were compared with experimental measurements. This report discusses various attributes of the rolled coupons revealed by the model (e.g., dog-boning and thickness non-uniformity).« less

Effects of the shape of the foil corners on the irradiation performance of U10Mo alloy based monolithic mini-plates

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

Ozaltun, Hakan; Medvedev, Pavel G

2015-06-01

Monolithic plate-type fuel is a fuel form being developed for high performance research and test reactors to minimize the use of enriched material. These fuel elements are comprised of a high density, low enrichment, U-Mo alloy based fuel foil, sandwiched between Zirconium liners and encapsulated in Aluminum cladding. The use of a high density fuel in a foil form presents a number of fabrication and operational concerns, such as: foil centering, flatness of the foil, fuel thickness variation, geometrical tilting, foil corner shape etc. To benchmark this new design, effects of various geometrical and operational variables on irradiation performance havemore » been evaluated. As a part of these series of sensitivity studies, the shape of the foil corners were studied. To understand the effects of the corner shapes of the foil on thermo-mechanical performance of the plates, a behavioral model was developed for a selected plate from RERTR-12 experiments (Plate L1P785). Both fabrication and irradiation processes were simulated. Once the thermo-mechanical behavior the plate is understood for the nominal case, the simulations were repeated for two additional corner shapes to observe the changes in temperature, displacement and stress-strain fields. The results from the fabrication simulations indicated that the foil corners do not alter the post-fabrication stress-strain magnitudes. Furthermore, the irradiation simulations revealed that post-fabrication stresses of the foil would be relieved very quickly in operation. While, foils with chamfered and filleted corners yielded stresses with comparable magnitudes, they are slightly lower in magnitudes, and provided a more favorable mechanical response compared with the foil with sharp corners.« less

Spent fuel reaction - the behavior of the {epsilon}-phase over 3.1 years

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

Finn, P.A.; Hoh, J.C.; Wolf, S.F.

The release fractions of the five elements in the {epsilon}-phase ({sup 99}Tc, {sup 97}Mo, Ru, Rh, and Pd) as well as that of {sup 238}U are reported for the reaction of two oxide fuels (ATM-103 and ATM-106) in unsaturated tests under oxidizing conditions. The {sup 99}Tc release fractions provide a lower limit for the magnitude of the spent fuel reaction. The {sup 99}Tc release fractions indicate that a surface reaction might be the rate controlling mechanism for fuel reaction under unsaturated conditions and the oxidant is possibly H{sub 2}O{sub 2}, a product of alpha radiolysis of water.

Antisense-based RNA therapy of factor V deficiency: in vitro and ex vivo rescue of a F5 deep-intronic splicing mutation.

PubMed

Nuzzo, Francesca; Radu, Claudia; Baralle, Marco; Spiezia, Luca; Hackeng, Tilman M; Simioni, Paolo; Castoldi, Elisabetta

2013-11-28

Antisense molecules are emerging as a powerful tool to correct splicing defects. Recently, we identified a homozygous deep-intronic mutation (F5 c.1296+268A>G) activating a cryptic donor splice site in a patient with severe coagulation factor V (FV) deficiency and life-threatening bleeding episodes. Here, we assessed the ability of 2 mutation-specific antisense molecules (a morpholino oligonucleotide [MO] and an engineered U7 small nuclear RNA [snRNA]) to correct this splicing defect. COS-1 and HepG2 cells transfected with a F5 minigene construct containing the patient's mutation expressed aberrant messenger RNA (mRNA) in excess of normal mRNA. Treatment with mutation-specific antisense MO (1-5 µM) or a construct expressing antisense U7snRNA (0.25-2 µg) dose-dependently increased the relative amount of correctly spliced mRNA by 1 to 2 orders of magnitude, whereas control MO and U7snRNA were ineffective. Patient-derived megakaryocytes obtained by differentiation of circulating progenitor cells did not express FV, but became positive for FV at immunofluorescence staining after administration of antisense MO or U7snRNA. However, treatment adversely affected cell viability, mainly because of the transfection reagents used to deliver the antisense molecules. Our data provide in vitro and ex vivo proof of principle for the efficacy of RNA therapy in severe FV deficiency, but additional cytotoxicity studies are warranted.

Elevated Temperature Tensile Tests on DU–10Mo Rolled Foils

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

Schulthess, Jason

2014-09-01

Tensile mechanical properties for uranium-10 wt.% molybdenum (U–10Mo) foils are required to support modeling and qualification of new monolithic fuel plate designs. It is expected that depleted uranium-10 wt% Mo (DU–10Mo) mechanical behavior is representative of the low enriched U–10Mo to be used in the actual fuel plates, therefore DU-10Mo was studied to simplify material processing, handling, and testing requirements. In this report, tensile testing of DU-10Mo fuel foils prepared using four different thermomechanical processing treatments were conducted to assess the impact of foil fabrication history on resultant tensile properties.

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

Besmann, Theodore M; Shin, Dongwon

TRISO coated particle fuel is envisioned as a next generation replacement for current urania pellet fuel in LWR applications. To obtain adequate fissile loading the kernel of the TRISO particle will need to be UN. In support of the fuel development effort, an assessment of phase regions of interest in the U-C-N system was undertaken as the fuel will be prepared by the carbothermic reduction of the oxide and it will be in equilibrium with carbon within the TRISO particle. The phase equilibria and thermochemistry of the U-C-N system is reviewed, including nitrogen pressure measurements above various phase fields. Selectedmore » measurements were used to fit a first order model of the UC1-xNx phase, represented by the inter-solution of UN and UC. Fit to the data was significantly improved by also adjusting the heat of formation for UN by ~12 kJ/mol and the phase equilbria was best reproduced by also adjusting the heat for U2N3 by +XXX. The determined interaction parameters yielded a slightly positive deviation from ideality, which agrees with lattice parameter measurements which show positive deviation from Vegard s law. The resultant model together with reported values for other phases in the system were used to generate isothermal sections of the U-C-N phase diagram. Nitrogen partial pressures were also computed for regions of interest.« less

AGR-5/6/7 Irradiation Test Predictions using PARFUME

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

Skerjanc, William F.

PARFUME, (PARticle FUel ModEl) a fuel performance modeling code used for high temperature gas-cooled reactors (HTGRs), was used to model the Advanced Gas Reactor (AGR)-5/6/7 irradiation test using predicted physics and thermal hydraulics data. The AGR-5/6/7 test consists of the combined fifth, sixth, and seventh planned irradiations of the AGR Fuel Development and Qualification Program. The AGR-5/6/7 test train is a multi-capsule, instrumented experiment that is designed for irradiation in the 133.4-mm diameter north east flux trap (NEFT) position of Advanced Test Reactor (ATR). Each capsule contains compacts filled with uranium oxycarbide (UCO) unaltered fuel particles. This report documents themore » calculations performed to predict the failure probability of tristructural isotropic (TRISO)-coated fuel particles during the AGR-5/6/7 experiment. In addition, this report documents the calculated source term from the driver fuel. The calculations include modeling of the AGR-5/6/7 irradiation that is scheduled to occur from October 2017 to April 2021 over a total of 13 ATR cycles, including nine normal cycles and four Power Axial Locator Mechanism (PALM) cycle for a total between 500 – 550 effective full power days (EFPD). The irradiation conditions and material properties of the AGR-5/6/7 test predicted zero fuel particle failures in Capsules 1, 2, and 4. Fuel particle failures were predicted in Capsule 3 due to internal particle pressure. These failures were predicted in the highest temperature compacts. Capsule 5 fuel particle failures were due to inner pyrolytic carbon (IPyC) cracking causing localized stresses concentrations in the SiC layer. This capsule predicted the highest particle failures due to the lower irradiation temperature. In addition, shrinkage of the buffer and IPyC layer during irradiation resulted in formation of a buffer-IPyC gap. The two capsules at the two ends of the test train, Capsules 1 and 5 experienced the smallest buffer-IPyC gap formation due to the lower irradiation fluences and temperatures. Capsule 3 experienced the largest buffer-IPyC gap formation of just under 24 µm. The release fraction of fission products Ag, Cs, and Sr silver (Ag), cesium (Cs), and strontium (Sr) vary depending on capsule location and irradiation temperature. The maximum release fraction of Ag occurs in Capsule 3, reaching up to 84.8% for the TRISO fuel particles. The release fraction of the other two fission products, Cs and Sr are much smaller and, in most cases, less than 1%. The notable exception is again in Capsule 3, where the release fraction for Cs and Sr reach up to 9.7% and 19.1%, respectively.« less

Chemical characterization of the fine particle emissions from commercial aircraft engines during the Aircraft Particle Emissions eXperiment (APEX) 1 to 3.

PubMed

Kinsey, J S; Hays, M D; Dong, Y; Williams, D C; Logan, R

2011-04-15

This paper addresses the need for detailed chemical information on the fine particulate matter (PM) generated by commercial aviation engines. The exhaust plumes of seven turbofan engine models were sampled as part of the three test campaigns of the Aircraft Particle Emissions eXperiment (APEX). In these experiments, continuous measurements of black carbon (BC) and particle surface-bound polycyclic aromatic compounds (PAHs) were conducted. In addition, time-integrated sampling was performed for bulk elemental composition, water-soluble ions, organic and elemental carbon (OC and EC), and trace semivolatile organic compounds (SVOCs). The continuous BC and PAH monitoring showed a characteristic U-shaped curve of the emission index (EI or mass of pollutant/mass of fuel burned) vs fuel flow for the turbofan engines tested. The time-integrated EIs for both elemental composition and water-soluble ions were heavily dominated by sulfur and SO(4)(2-), respectively, with a ∼2.4% median conversion of fuel S(IV) to particle S(VI). The corrected OC and EC emission indices obtained in this study ranged from 37 to 83 mg/kg and 21 to 275 mg/kg, respectively, with the EC/OC ratio ranging from ∼0.3 to 7 depending on engine type and test conditions. Finally, the particle SVOC EIs varied by as much as 2 orders of magnitude with distinct variations in chemical composition observed for different engine types and operating conditions.

Radioactive particles released to the environment from the Fukushima reactors-Confirmation is still needed.

PubMed

Salbu, Brit; Lind, Ole Christian

2016-10-01

After severe nuclear events, a major fraction of refractory radionuclides such as U and Pu are released to the environment in the form of radioactive particles. After the Fukushima Daiichi Nuclear Power Plant (FDNPP) accident, Pu isotope ratio signals different from that of global fallout have been reported, indicating that spent fuel particles have been released from the reactors or reactor vessels. Radioactive particles containing (37) Cs and other volatile radionuclides, as well as a series of stable refractory metals (Cs, Fe, Zn, U, etc.), have been identified by several authors claiming that these particles originated from the FDNPP fuel. If so, long-lived radioactive isotopes of the refractory metals should have been identified in these particles. It is therefore most probable that volatile radionuclides released as gases during the accidents have deposited on available surfaces such as fly ash, forming condensation particles during release or transport. If spent fuel particles have been deposited in the FDNPP surroundings, information on particle characteristics influencing ecosystem transport, uptake, and effects is essential for assessing environmental impact and risk. More emphasis should therefore be put on the identification of hot spots in the FDNPP environment followed by the characterization of radioactive particles using nanoanalytical-microanalytical techniques to support environmental monitoring, as recommended in the present study. Integr Environ Assess Manag 2016;12:687-689. © 2016 SETAC. © 2016 SETAC.

Characterization of SrFe0.75Mo0.25O3-δ-La0.9Sr0.1Ga0.8Mg0.2O3-δ composite cathodes prepared by infiltration

NASA Astrophysics Data System (ADS)

Meng, Xie; Han, Da; Wu, Hao; Li, Junliang; Zhan, Zhongliang

2014-01-01

This paper describes the structure and electrochemical properties of composite cathodes for solid oxide fuel cells fabricated by infiltration of aqueous solutions corresponding to SrFe0.75Mo0.25O3-δ (SFMO) into porous La0.9Sr0.1Ga0.8Mg0.2O3-δ (LSGM) backbones. XRD measurement confirms the predominance of the perovskite SFMO oxides in the infiltrates together with some minor impurities of SrMoO4 after calcinations at 850-1100 °C. The cathode polarization resistance as obtained from impedance measurement on symmetric cathode fuel cells exhibits a pronounced increase as a function of calcinations temperature due to increased SFMO particle sizes, e.g., 0.04 Ω cm2 for 70 nm-sized catalysts calcinated at 850 °C versus 0.11 Ω cm2 for 400 nm-sized catalysts calcinated at 1100 °C. Oxygen partial pressure and temperature dependence of impedance data shows that oxygen reduction kinetics is largely determined by ionization of adsorbed oxygen atoms on the SFMO catalysts.

Grain boundary engineering to control the discontinuous precipitation in multicomponent U10Mo alloy

DOE PAGES

Devaraj, Arun; Kovarik, Libor; Kautz, Elizabeth; ...

2018-03-30

Here, we demonstrate here that locally stabilized structure and compositional segregation at grain boundaries in a complex multicomponent alloy can be modified using high temperature homogenization treatment to influence the kinetics of phase transformations initiating from grain boundaries during subsequent low temperature annealing. Using aberration-corrected scanning transmission electron microscopy and atom probe tomography of a model multicomponent metallic alloy —uranium-10 wt% molybdenum (U-10Mo) a nuclear fuel, that is highly relevant to worldwide nuclear non-proliferation efforts, we demonstrate the ability to change the structure and compositional segregation at grain boundary, which then controls the subsequent discontinuous precipitation kinetics during sub-eutectoid annealing.more » A change in grain boundary from one characterized by segregation of Mo and impurities at grain boundary to a phase boundary with a distinct U 2MoSi 2C wetting phase precipitates introducing Ni and Al rich interphase complexions caused a pronounced reduction in area fraction of subsequent discontinuous precipitation. The broader implication of this work is in highlighting the role of grain boundary structure and composition in metallic alloys on dictating the fate of grain boundary initiated phase transformations like discontinuous precipitation or cellular transformation. This work highlights a new pathway to tune the grain boundary structure and composition to tailor the final microstructure of multicomponent metallic alloys.« less

Grain boundary engineering to control the discontinuous precipitation in multicomponent U10Mo alloy

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

Devaraj, Arun; Kovarik, Libor; Kautz, Elizabeth

Here, we demonstrate here that locally stabilized structure and compositional segregation at grain boundaries in a complex multicomponent alloy can be modified using high temperature homogenization treatment to influence the kinetics of phase transformations initiating from grain boundaries during subsequent low temperature annealing. Using aberration-corrected scanning transmission electron microscopy and atom probe tomography of a model multicomponent metallic alloy —uranium-10 wt% molybdenum (U-10Mo) a nuclear fuel, that is highly relevant to worldwide nuclear non-proliferation efforts, we demonstrate the ability to change the structure and compositional segregation at grain boundary, which then controls the subsequent discontinuous precipitation kinetics during sub-eutectoid annealing.more » A change in grain boundary from one characterized by segregation of Mo and impurities at grain boundary to a phase boundary with a distinct U 2MoSi 2C wetting phase precipitates introducing Ni and Al rich interphase complexions caused a pronounced reduction in area fraction of subsequent discontinuous precipitation. The broader implication of this work is in highlighting the role of grain boundary structure and composition in metallic alloys on dictating the fate of grain boundary initiated phase transformations like discontinuous precipitation or cellular transformation. This work highlights a new pathway to tune the grain boundary structure and composition to tailor the final microstructure of multicomponent metallic alloys.« less

Screening of advanced cladding materials and UN-U3Si5 fuel

NASA Astrophysics Data System (ADS)

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

2015-07-01

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

U-10Mo Baseline Fuel Fabrication Process Description

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

Hubbard, Lance R.; Arendt, Christina L.; Dye, Daniel F.

This document provides a description of the U.S. High Power Research Reactor (USHPRR) low-enriched uranium (LEU) fuel fabrication process. This document is intended to be used in conjunction with the baseline process flow diagram (PFD) presented in Appendix A. The baseline PFD is used to document the fabrication process, communicate gaps in technology or manufacturing capabilities, convey alternatives under consideration, and as the basis for a dynamic simulation model of the fabrication process. The simulation model allows for the assessment of production rates, costs, and manufacturing requirements (manpower, fabrication space, numbers and types of equipment, etc.) throughout the lifecycle ofmore » the USHPRR program. This document, along with the accompanying PFD, is updated regularly« less

Molybdenum dioxide-based anode for solid oxide fuel cell applications

NASA Astrophysics Data System (ADS)

Kwon, Byeong Wan; Ellefson, Caleb; Breit, Joe; Kim, Jinsoo; Grant Norton, M.; Ha, Su

2013-12-01

The present paper describes the fabrication and performance of a molybdenum dioxide (MoO2)-based anode for liquid hydrocarbon/oxygenated hydrocarbon-fueled solid oxide fuel cells (SOFCs). These fuel cells first internally reform the complex liquid fuel into carbon fragments and hydrogen, which are then electrochemically oxidized to produce electrical energy without external fuel processors. The MoO2-based anode was fabricated on to an yttria-stabilized zirconia (YSZ) electrolyte via combined electrostatic spray deposition (ESD) and direct painting methods. The cell performance was measured by directly feeding liquid fuels such as n-dodecane (i.e., a model diesel/kerosene fuel) or biodiesel (i.e., a future biomass-based liquid fuel) to the MoO2-based anode at 850 °C. The maximum initial power densities obtained from our MoO2-based SOFC were 34 mW cm-2 and 45 mW cm-2 using n-dodecane and biodiesel, respectively. The initial power density of the MoO2-based SOFC was improved up to 2500 mW cm-2 by optimizing the porosity of the MoO2-based anode. To test the long-term stability of the MoO2-based anode SOFC against coking, n-dodecane was continuously fed into the cell for 24 h at the open circuit voltage (OCV). During long-term testing, voltage-current density (V-I) plots were periodically obtained and they showed no significant changes over the operation time. Microstructural examination of the tested cells indicated that the MoO2-based anode displayed negligible coke formation, which explains its stability. On the other hand, SOFCs with conventional nickel (Ni)-based anodes under the same operating conditions showed a significant amount of coke formation on the metal surface, which led to a rapid drop in cell performance. Hence, the present work demonstrates that MoO2-based anodes exhibit outstanding tolerance to coke formation. This result opens up the opportunity for more efficiently generating electrical energy from both existing transportation and next generation biomass-derived liquid fuels using liquid hydrocarbon/oxygenated hydrocarbon-fueled SOFCs.

Synthesis and Characterization of CO- and H2S-Tolerant Electrocatalysts for PEM Fuel Cell

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

Shamsuddin Ilias

2006-05-18

The present state-of-art Proton Exchange Membrane Fuel Cell (PEMFC) technology is based on platinum (Pt) as a catalyst for both the fuel (anode) and air (cathode) electrodes. This catalyst is highly active but susceptible to poisoning by CO, which may be present in the H{sub 2}-fuel used or may be introduced during the fuel processing. Presence of trace amount of CO and H{sub 2}S in the H{sub 2}-fuel poisons the anode irreversibly and decreases the performance of the PEMFCs. In an effort to reduce the Pt-loading and improve the PEMFC performance, we propose to synthesize a number of Pt-based binary,more » ternary, and quaternary electrocatalysts using Ru, Mo, Ir, Ni, and Co as a substitute for Pt. By fine-tuning the metal loadings and compositions of candidate electrocatalysts, we plan to minimize the cost and optimize the catalyst activity and performance in PEMFC. The feasibility of the novel electrocatalysts will be demonstrated in the proposed effort with gas phase CO and H{sub 2}S concentrations typical of those found in reformed fuel gas with coal/natural gas/methanol feedstocks. During this reporting period we used four Pt-based electrocatalysts (Pt/Ru/Mo/Se, Pt/Ru/Mo/Ir, Pt/Ru/Mo/W, Ptr/Ru/Mo/Co) in MEAs and these were evaluated for CO-tolerance with 20 and 100 ppm CO concentration in H{sub 2}-fuel. From current-voltage performance study, the catalytic activity was found in the increasing order of Pt/Ru/Mo/Ir > Pt/Ru/Mo/W > Pt/Ru/Mo/Co > Pt/Ru/MO/Se. From preliminary cost analysis it appears that could of the catalyst metal loading can reduced by 40% to 60% depending on the selection of metal combinations without compromising the fuel cell performance.« less

Magnetic MoS2 on multiwalled carbon nanotubes for sulfide sensing.

PubMed

Li, Chunxiang; Zhang, Dan; Wang, Jiankang; Hu, Pingan; Jiang, Zhaohua

2017-07-04

A novel hybrid metallic cobalt insided in multiwalled carbon nanotubles/molybdenum disulfide (Co@CNT/MoS 2 ) modified glass carbon electrode (GCE) was fabricated with a adhesive of Nafion suspension and used as chemical sensors for sulfide detection. Single-layered MoS 2 was coated on CNTs through magnetic traction force between paramagnetic monolayer MoS 2 and Co particles in CNTs. Co particles faciliated the collection of paramagnetic monolayer MoS 2 exfoliated from bulk MoS 2 in solution. Amperometric analysis, cycle voltammetry, cathodic stripping analysis and linear sweep voltammetry results showed the Co@CNT/MoS 2 modified GCE exhibited excellent electrochemical activity to sulfide in buffer solutions, but amperometric analysis was found to be more sensitive than the other methods. The amperometric response result indicated the Co@CNT/MoS 2 -modified GCE electrode was an excellent electrochemical sensor for detecting S 2- with a detection limit of 7.6 nM and sensitivity of 0.23 mA/μM. The proposed electrode was used for the determination of sulfide levels in hydrogen sulfide-pretreated fruits, and the method was also verified with recovery studies. Copyright © 2017 Elsevier B.V. All rights reserved.

76 FR 40765 - Missouri Disaster Number MO-00048

Federal Register 2010, 2011, 2012, 2013, 2014

2011-07-11

... SMALL BUSINESS ADMINISTRATION [Disaster Declaration 12576 and 12577] Missouri Disaster Number MO-00048 AGENCY: U.S. Small Business Administration. ACTION: Amendment 7. SUMMARY: This is an amendment of... information in the original declaration remains unchanged. (Catalog of Federal Domestic Assistance Numbers...

Nanoparticle Precipitation in Irradiated and Annealed Ceria Doped with Metals for Emulation of Spent Fuels

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

Jiang, Weilin; Conroy, Michele A.; Kruska, Karen

Epsilon alloy precipitates have been observed with varied compositions and sizes in spent nuclear fuels, such as UO2. Presence of the inclusions, along with other oxide precipitates, gas bubbles and irradiation-induced structural defects, can significantly degrade the physical properties of the fuel. To predict fuel performance, a fundamental study of the precipitation processes is needed. This study uses ceria (CeO2) as a surrogate for UO2. Polycrystalline CeO2 films doped with Mo, Ru, Rh, Pd and Re (surrogate for Tc) were grown at 823 K using pulsed laser deposition, irradiated at 673 K with He+ ions, and subsequently annealed at highermore » temperatures. A number of methods, including transmission electron microscopy and atom probe tomography, were applied to characterize the samples. The results indicate that there is a uniform distribution of the doped metals in the as-grown CeO2 film. Pd particles of ~3 nm in size appear near dislocation edges after He+ ion irradiation to ~13 dpa. Thermal annealing at 1073 K in air leads to formation of precipitates with Mo and Pd around grain boundaries. Further annealing at 1373 K produces 70 nm sized precipitates with small grains at cavities.« less

USHPRR FUEL FABRICATION PILLAR: FABRICATION STATUS, PROCESS OPTIMIZATIONS, AND FUTURE PLANS

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

Wight, Jared M.; Joshi, Vineet V.; Lavender, Curt A.

The Fuel Fabrication (FF) Pillar, a project within the U.S. High Performance Research Reactor Conversion program of the National Nuclear Security Administration’s Office of Material Management and Minimization, is tasked with the scale-up and commercialization of high-density monolithic U-Mo fuel for the conversion of appropriate research reactors to use of low-enriched fuel. The FF Pillar has made significant steps to demonstrate and optimize the baseline co-rolling process using commercial-scale equipment at both the Y-12 National Security Complex (Y-12) and BWX Technologies (BWXT). These demonstrations include the fabrication of the next irradiation experiment, Mini-Plate 1 (MP-1), and casting optimizations at Y-12.more » The FF Pillar uses a detailed process flow diagram to identify potential gaps in processing knowledge or demonstration, which helps direct the strategic research agenda of the FF Pillar. This paper describes the significant progress made toward understanding the fuel characteristics, and models developed to make informed decisions, increase process yield, and decrease lifecycle waste and costs.« less

Effects of the foil flatness on irradiation performance of U10Mo monolithic mini-plates

DOE PAGES

Ozaltun, Hakan; Medvedev, Pavel G.; Rabin, Barry H.

2015-09-03

Monolithic plate-type fuels comprise of a high density, low enrichment, U10Mo fuel foil encapsulated in a cladding material. This concept generates several fabrication challenges such as flatness, centering or thickness variation. There are concerns, if these parameters have implications on overall performance. To investigate these inquiries, the effects of the foil flatness were studied. For this, a representative plate was simulated for an ideal case. The simulations were repeated for additional cases with various foil curvatures to evaluate the effects on the irradiation performance. The results revealed that the stresses and strains induced by fabrication process are not affected bymore » the flatness of the foil. Furthermore, fabrication stresses in the foil are relieved relatively fast in the reactor. The effects of the foil flatness on peak irradiation stressstrains are minimal. There is a slight increase in temperature for the case with maximum curvature. The major impact is on the displacement characteristics. Furthermore, while the case with a flat foil produces a symmetrical swelling, if the foil is curved, more swelling occurs on the thin-cladding side and the plate bows during irradiation.« less

Impact of thorium based molten salt reactor on the closure of the nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Jaradat, Safwan Qasim Mohammad

Molten salt reactor (MSR) is one of six reactors selected by the Generation IV International Forum (GIF). The liquid fluoride thorium reactor (LFTR) is a MSR concept based on thorium fuel cycle. LFTR uses liquid fluoride salts as a nuclear fuel. It uses 232Th and 233U as the fertile and fissile materials, respectively. Fluoride salt of these nuclides is dissolved in a mixed carrier salt of lithium and beryllium (FLiBe). The objective of this research was to complete feasibility studies of a small commercial thermal LFTR. The focus was on neutronic calculations in order to prescribe core design parameter such as core size, fuel block pitch (p), fuel channel radius, fuel path, reflector thickness, fuel salt composition, and power. In order to achieve this objective, the applicability of Monte Carlo N-Particle Transport Code (MCNP) to MSR modeling was verified. Then, a prescription for conceptual small thermal reactor LFTR and relevant calculations were performed using MCNP to determine the main neutronic parameters of the core reactor. The MCNP code was used to study the reactor physics characteristics for the FUJI-U3 reactor. The results were then compared with the results obtained from the original FUJI-U3 using the reactor physics code SRAC95 and the burnup analysis code ORIPHY2. The results were comparable with each other. Based on the results, MCNP was found to be a reliable code to model a small thermal LFTR and study all the related reactor physics characteristics. The results of this study were promising and successful in demonstrating a prefatory small commercial LFTR design. The outcome of using a small core reactor with a diameter/height of 280/260 cm that would operate for more than five years at a power level of 150 MWth was studied. The fuel system 7LiF - BeF2 - ThF4 - UF4 with a (233U/ 232Th) = 2.01 % was the candidate fuel for this reactor core.

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).

Uranium Dioxides and Debris Fragments Released to the Environment with Cesium-Rich Microparticles from the Fukushima Daiichi Nuclear Power Plant.

PubMed

Ochiai, Asumi; Imoto, Junpei; Suetake, Mizuki; Komiya, Tatsuki; Furuki, Genki; Ikehara, Ryohei; Yamasaki, Shinya; Law, Gareth T W; Ohnuki, Toshihiko; Grambow, Bernd; Ewing, Rodney C; Utsunomiya, Satoshi

2018-03-06

Trace U was released from the Fukushima Daiichi Nuclear Power Plant (FDNPP) during the meltdowns, but the speciation of the released components of the nuclear fuel remains unknown. We report, for the first time, the atomic-scale characteristics of nanofragments of the nuclear fuels that were released from the FDNPP into the environment. Nanofragments of an intrinsic U-phase were discovered to be closely associated with radioactive cesium-rich microparticles (CsMPs) in paddy soils collected ∼4 km from the FDNPP. The nanoscale fuel fragments were either encapsulated by or attached to CsMPs and occurred in two different forms: (i) UO 2+X nanocrystals of ∼70 nm size, which are embedded into magnetite associated with Tc and Mo on the surface and (ii) Isometric (U,Zr)O 2+X nanocrystals of ∼200 nm size, with the U/(U+Zr) molar ratio ranging from 0.14 to 0.91, with intrinsic pores (∼6 nm), indicating the entrapment of vapors or fission-product gases during crystallization. These results document the heterogeneous physical and chemical properties of debris at the nanoscale, which is a mixture of melted fuel and reactor materials, reflecting the complex thermal processes within the FDNPP reactor during meltdown. Still CsMPs are an important medium for the transport of debris fragments into the environment in a respirable form.

Synthesis and Characterization of CO-and H2S-Tolerant Electrocatalysts for PEM Fuel Cell

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

Shamsuddin Ilias

2005-12-22

The present state-of-art Proton Exchange Membrane Fuel Cell (PEMFC) technology is based on platinum (Pt) as a catalyst for both the fuel (anode) and air (cathode) electrodes. This catalyst is highly active but susceptible to poisoning by CO, which may be present in the H{sub 2}-fuel used or may be introduced during the fuel processing. Presence of trace amount of CO and H{sub 2}S in the H{sub 2}-fuel poisons the anode irreversibly and decreases the performance of the PEMFCs. In an effort to reduce the Pt-loading and improve the PEMFC performance, we propose to synthesize a number of Pt-based binary,more » ternary, and quaternary electrocatalysts using Ru, Mo, Ir, Ni, and Co as a substitute for Pt. By fine-tuning the metal loadings and compositions of candidate electrocatalysts, we plan to minimize the cost and optimize the catalyst activity and performance in PEMFC. The feasibility of the novel electrocatalysts will be demonstrated in the proposed effort with gas phase CO and H{sub 2}S concentrations typical of those found in reformed fuel gas with coal/natural gas/methanol feedstocks. During this reporting period we synthesized four Pt-based electrocatalysts catalysts (Pt/Ru/Mo/Se, Pt/Ru/Mo/Ir, Pt/Ru/Mo/W, Ptr/Ru/Mo/Co) on Vulcan XG72 Carbon support by both conventional and ultra-sonication method. From current-voltage performance study, the catalytic activity was found in the increasing order of Pt/Ru/Mo/Ir > Pt/Ru/Mo/W > Pt/Ru/Mo/Co > Pt/Ru/MO/Se. Sonication method appears to provide better dispersion of catalysts on carbon support.« less

AGR-3/4 Irradiation Test Predictions using PARFUME

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

Skerjanc, William Frances; Collin, Blaise Paul

2016-03-01

PARFUME, a fuel performance modeling code used for high temperature gas reactors, was used to model the AGR-3/4 irradiation test using as-run physics and thermal hydraulics data. The AGR-3/4 test is the combined third and fourth planned irradiations of the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The AGR-3/4 test train consists of twelve separate and independently controlled and monitored capsules. Each capsule contains four compacts filled with both uranium oxycarbide (UCO) unaltered “driver” fuel particles and UCO designed-to-fail (DTF) fuel particles. The DTF fraction was specified to be 1×10-2. This report documents the calculations performed to predictmore » failure probability of TRISO-coated fuel particles during the AGR-3/4 experiment. In addition, this report documents the calculated source term from both the driver fuel and DTF particles. The calculations include the modeling of the AGR-3/4 irradiation that occurred from December 2011 to April 2014 in the Advanced Test Reactor (ATR) over a total of ten ATR cycles including seven normal cycles, one low power cycle, one unplanned outage cycle, and one Power Axial Locator Mechanism cycle. Results show that failure probabilities are predicted to be low, resulting in zero fuel particle failures per capsule. The primary fuel particle failure mechanism occurred as a result of localized stresses induced by the calculated IPyC cracking. Assuming 1,872 driver fuel particles per compact, failure probability calculated by PARFUME leads to no predicted particle failure in the AGR-3/4 driver fuel. In addition, the release fraction of fission products Ag, Cs, and Sr were calculated to vary depending on capsule location and irradiation temperature. The maximum release fraction of Ag occurs in Capsule 7 reaching up to 56% for the driver fuel and 100% for the DTF fuel. The release fraction of the other two fission products, Cs and Sr, are much smaller and in most cases less than 1% for the driver fuel. The notable exception occurs in Capsule 7 where the release fraction for Cs and Sr reach up to 0.73% and 2.4%, respectively, for the driver fuel. For the DTF fuel in Capsule 7, the release fraction for Cs and Sr are estimated to be 100% and 5%, respectively.« less

Dynamic leaching studies of 48 MWd/kgU UO2 commercial spent nuclear fuel under oxic conditions

NASA Astrophysics Data System (ADS)

Serrano-Purroy, D.; Casas, I.; González-Robles, E.; Glatz, J. P.; Wegen, D. H.; Clarens, F.; Giménez, J.; de Pablo, J.; Martínez-Esparza, A.

2013-03-01

The leaching of a high-burn-up spent nuclear fuel (48 MWd/KgU) has been studied in a carbonate-containing solution and under oxic conditions using a Continuously Stirred Tank Flow-Through Reactor (CSTR). Two samples of the fuel, one prepared from the centre of the pellet (labelled CORE) and another one from the fuel pellet periphery, enriched with the so-called High Burn-Up Structure (HBS, labelled OUT) have been used.For uranium and actinides, the results showed that U, Np, Am and Cm gave very similar normalized dissolution rates, while Pu showed slower dissolution rates for both samples. In addition, dissolution rates were consistently two to four times lower for OUT sample compared to CORE sample.Considering the fission products release the main results are that Y, Tc, La and Nd dissolved very similar to uranium; while Cs, Sr, Mo and Rb have up to 10 times higher dissolution rates. Rh, Ru and Zr seemed to have lower dissolution rates than uranium. The lowest dissolution rates were found for OUT sample.Three different contributions were detected on uranium release, modelled and attributed to oxidation layer, fines and matrix release.

Enhanced hydrogen generation by hydrolysis of Mg doped with flower-like MoS2 for fuel cell applications

NASA Astrophysics Data System (ADS)

Huang, Minghong; Ouyang, Liuzhang; Liu, Jiangwen; Wang, Hui; Shao, Huaiyu; Zhu, Min

2017-10-01

In this work, flower-like MoS2 spheres are synthesized via a hydrothermal method and the catalytic activity of the as-prepared and bulk MoS2 on hydrolysis of Mg is systematically investigated for the first time. The Mg-MoS2 composites are prepared by ball milling and the hydrogen generation performances of the composites are investigated in 3.5% NaCl solution. The experimental results suggest that the as-prepared MoS2 exhibits better catalytic effect on hydrolysis of Mg compared to bulk MoS2. In particular, Mg-10 wt% MoS2 (as-prepared) composite milled for 1 h shows the best hydrogen generation properties and releases 90.4% of theoretical hydrogen generation capacity within 1 min at room temperature. The excellent catalytic effect of as-prepared MoS2 may be attributed to the following aspects: three-dimensional flower-like MoS2 architectures improve its dispersibility on Mg particles; make the composite more reactive; hamper the generated Mg(OH)2 from adhering to the surface of Mg; and increase the galvanic corrosion of Mg. In addition, a hydrogen generator based on the hydrolysis reaction of Mg-0.2 wt% MoS2 composite is manufactured and it can supply a maximum hydrogen flow rate of 2.5 L/min. The findings here demonstrate the as-prepared flower-like MoS2 can be a promising catalyst for hydrogen generation from Mg.

Nano-molybdenum carbide/carbon nanotubes composite as bifunctional anode catalyst for high-performance Escherichia coli-based microbial fuel cell.

PubMed

Wang, Yaqiong; Li, Bin; Cui, Dan; Xiang, Xingde; Li, Weishan

2014-01-15

A novel electrode, carbon felt-supported nano-molybdenum carbide (Mo2C)/carbon nanotubes (CNTs) composite, was developed as platinum-free anode of high performance microbial fuel cell (MFC). The Mo2C/CNTs composite was synthesized by using the microwave-assisted method with Mo(CO)6 as a single source precursor and characterized by using X-ray diffraction and transmission electron microscopy. The activity of the composite as anode electrocatalyst of MFC based on Escherichia coli (E. coli) was investigated with cyclic voltammetry, chronoamperometry, and cell discharge test. It is found that the carbon felt electrode with 16.7 wt% Mo Mo2C/CNTs composite exhibits a comparable electrocatalytic activity to that with 20 wt% platinum as anode electrocatalyst. The superior performance of the developed platinum-free electrode can be ascribed to the bifunctional electrocatalysis of Mo2C/CNTs for the conversion of organic substrates into electricity through bacteria. The composite facilitates the formation of biofilm, which is necessary for the electron transfer via c-type cytochrome and nanowires. On the other hand, the composite exhibits the electrocatalytic activity towards the oxidation of hydrogen, which is the common metabolite of E. coli. © 2013 Elsevier B.V. All rights reserved.

Laboratory characterization of PM emissions from combustion of wildland biomass fuels

Treesearch

Seyedehsan Hosseini; Shawn Urbanski; P. Dixit; Qi Li; Ian Burling; Robert Yokelson; Timothy E. Johnson; Manish Sharivastava; Heejung Jung; David R. Weise; Wayne Miller; David Cocker

2013-01-01

Particle emissions from open burning of southwestern (SW) and southeastern (SE) U.S. fuel types during 77 controlled laboratory burns are presented. The fuels include SW vegetation types: ceanothus, chamise/scrub oak, coastal sage scrub, California sagebrush, manzanita, maritime chaparral, masticated mesquite, oak savanna, and oak woodland, as well as SE vegetation...

Effects of High Availability Fuels on Combustor Properties

DTIC Science & Technology

1978-01-01

EFFECT OF HIGH AVAILABILITY FUELS ON COMBUSTOR PROPERTIES INTERIM REPORT AFLRL No. 101 by C. A. Moses and D. W. Naegeli prepared by U. S. Army Fuels...C-,0•)a3 D.W. / Naege I i_• //• DAAK 7/,0’ 7 8 - C -,9ak1 9 PERFORMING ORGANIZATION NAME AND ADDRESSES - ,_ _ECT, TASK U.S. Army Fuels & Lubricants

Performance comparison of protonic and sodium phosphomolybdovanadate polyoxoanion catholytes within a chemically regenerative redox cathode polymer electrolyte fuel cell

NASA Astrophysics Data System (ADS)

Ward, David B.; Gunn, Natasha L. O.; Uwigena, Nadine; Davies, Trevor J.

2018-01-01

The direct reduction of oxygen in conventional polymer electrolyte fuel cells (PEFCs) is seen by many researchers as a key challenge in PEFC development. Chemically regenerative redox cathode (CRRC) polymer electrolyte fuel cells offer an alternative approach via the indirect reduction of oxygen, improving durability and reducing cost. These systems substitute gaseous oxygen for a liquid catalyst that is reduced at the cathode then oxidised in a regeneration vessel via air bubbling. A key component of a CRRC system is the liquid catalyst or catholyte. To date, phosphomolybdovanadium polyoxometalates with empirical formula H3+nPVnMo12-nO40 have shown the most promise for CRRC PEFC systems. In this work, four catholyte formulations are studied and compared against each other. The catholytes vary in vanadium content, pH and counter ion, with empirical formulas H6PV3Mo9O40, H7PV4Mo8O40, Na3H3PV3Mo9O40 and Na4H3PV4Mo8O40. Thermodynamic properties, cell performance and regeneration rates are measured, generating new insights into how formulation chemistry affects the components of a CRRC system. The results include the best CRRC PEFC performance reported to date, with noticeable advantages over conventional PEFCs. The optimum catholyte formulation is then determined via steady state tests, the results of which will guide further optimization of the catholyte formulation.

Assessment of Nuclear Fuels using Radiographic Thickness Measurement Method

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

Muhammad Abir; Fahima Islam; Hyoung Koo Lee

2014-11-01

The Convert branch of the National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI) focuses on the development of high uranium density fuels for research and test reactors for nonproliferation. This fuel is aimed to convert low density high enriched uranium (HEU) based fuel to high density low enriched uranium (LEU) based fuel for high performance research reactors (HPRR). There are five U.S. reactors that fall under the HPRR category, including: the Massachusetts Institute of Technology Reactor (MITR), the National Bureau of Standards Reactor (NBSR), the Missouri University Research Reactor (UMRR), the Advanced Test Reactor (ATR), and the Highmore » Flux Isotope Reactor (HFIR). U-Mo alloy fuel phase in the form of either monolithic or dispersion foil type fuels, such as ATR Full-size In center flux trap Position (AFIP) and Reduced Enrichment for Research and Test Reactor (RERTR), are being designed for this purpose. The fabrication process1 of RERTR is susceptible to introducing a variety of fuel defects. A dependable quality control method is required during fabrication of RERTR miniplates to maintain the allowable design tolerances, therefore evaluating and analytically verifying the fabricated miniplates for maintaining quality standards as well as safety. The purpose of this work is to analyze the thickness of the fabricated RERTR-12 miniplates using non-destructive technique to meet the fuel plate specification for RERTR fuel to be used in the ATR.« less

FastDart : a fast, accurate and friendly version of DART code.

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

Rest, J.; Taboada, H.

2000-11-08

A new enhanced, visual version of DART code is presented. DART is a mechanistic model based code, developed for the performance calculation and assessment of aluminum dispersion fuel. Major issues of this new version are the development of a new, time saving calculation routine, able to be run on PC, a friendly visual input interface and a plotting facility. This version, available for silicide and U-Mo fuels,adds to the classical accuracy of DART models for fuel performance prediction, a faster execution and visual interfaces. It is part of a collaboration agreement between ANL and CNEA in the area of Lowmore » Enriched Uranium Advanced Fuels, held by the Implementation Arrangement for Technical Exchange and Cooperation in the Area of Peaceful Uses of Nuclear Energy.« less

Bioaccessibility of micron-sized powder particles of molybdenum metal, iron metal, molybdenum oxides and ferromolybdenum--Importance of surface oxides.

PubMed

Mörsdorf, Alexander; Odnevall Wallinder, Inger; Hedberg, Yolanda

2015-08-01

The European chemical framework REACH requires that hazards and risks posed by chemicals, including alloys and metals, that are manufactured, imported or used in different products (substances or articles) are identified and proven safe for humans and the environment. Metals and alloys need hence to be investigated on their extent of released metals (bioaccessibility) in biologically relevant environments. Read-across from available studies may be used for similar materials. This study investigates the release of molybdenum and iron from powder particles of molybdenum metal (Mo), a ferromolybdenum alloy (FeMo), an iron metal powder (Fe), MoO2, and MoO3 in different synthetic body fluids of pH ranging from 1.5 to 7.4 and of different composition. Spectroscopic tools and cyclic voltammetry have been employed to characterize surface oxides, microscopy, light scattering and nitrogen absorption for particle characterization, and atomic absorption spectroscopy to quantify released amounts of metals. The release of molybdenum from the Mo powder generally increased with pH and was influenced by the fluid composition. The mixed iron and molybdenum surface oxide of the FeMo powder acted as a barrier both at acidic and weakly alkaline conditions. These findings underline the importance of the surface oxide characteristics for the bioaccessibility of metal alloys. Copyright © 2015 The Authors. Published by Elsevier Inc. All rights reserved.

METALLURGY DIVISION QUARTERLY REPORT FOR JULY, AUGUST, AND SEPTEMBER 1957

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

None

1958-10-01

Advanced Water Reactor Program. Three firings were made of initial closed-porosity fuel pellet bodies. Each firing coatained pellets of the composition 90 wt.% ThO/sub 2/-10 wt.%fl U0/sub 2/ with various additives and firing variables. Fast Power Breeder Reactor Program. To determine the potential usefulness of a Zr-5 wt. % Pu alloy, the fabricability of the alloy was tested. The manufacture of rod stock from which fuel and blanket elements for the Mark III loading of the EBR-1 were prcduced has been completed. The effect of irradiation on extruded and heat-treated U-2 wt.% Zr alloy for the EBR- 1 is reported.more » Fabrication procedures for making graphite-U/sub 3/O/sub 8/ test specimens for the TREAT Reactor were investigated. Advanced Engineering and Development. Ultrasonic bond tests were conducted on 590 EBR-1 Mark III blanket fuel elemeats. The blanket rods and part of the fuel rcds for the EBR-1 Mark III loading are being checked for cladding thickness by an eddy current system. Investigations of corrosionresistant Zr-Nb alloy were coatinued. Corrosion of MR alloys is being studied Ln support of the Mighty Mouse reactor program. Dynamic corrosion tests were performed on aluminum alloys, and results are included. Prcduction, Treatment, and Properties of Materials. The progress of the program of preparing highpurity Pu by fused salt electrolysis is summarized. Velocities of ultrasonic waves propagated in directions suitable for determining the room- temperature elastic moduli C/sub 12/, C/sub 13/, and C/sub 23/ of alpha U were determined. investigation of recrystallization in heavily coldrolled alpha- uranium sheet without a texture change was essentially concluded during this quarter. Selfdiffasion runs in polycrystalline uranium in the gamma phase, using the sputtering technique, have yielded a tentative value for the diffusion coefficient between 10/sup -8/ and 10/sup -7/ cm/sup 2/second. The preparation of high-purity U-Pan alloys is reponted. The data for the alpha-tobeta transformation temperatures in high-purity U and U-C alloys were confirmed by experiments on new specimens. Microstructure, density, and thermal arrest data were obtained for several injection cast, nominal U-5 wt.%fl fissium and U-8 wt.%fl fissium alloys. Phase diagrams are preseated for U-Mo and U-Ru alloys. Alloy Theory and The Nature of Solids. Four new isomorphs of Ti/sub 2/Ni have been discovered. Effects of Irradiation on Materials. The experimental and analytical work on the radial distribution of thermal neutrons within cylindrically shaped fuel specimens during irradiation was completed. (For preceding period see ANL-5790.) (W.L.H.)« less

Properties of Aircraft Fuels and Related Materials.

DTIC Science & Technology

1991-07-29

LA . 0I Z La u0I Z4 Li LA . L U) a:L) mi F- - L Wz N ) Mtwm Li La . N t o) a f) r en f" ’m, Mww 4v Nv NuQe...mo 000) 0 100 U4v4U1 w 70 w n(~xI I ! I- I I iI I i I I I I Ii I I I " LJLaJ La . LaJLI 0 0 L.. ao . 0 .L 0 EL 17z U. Z IV CD 0 Iqx 0 1 O0 cN.. 10. In...notice on a specific document. INCLASSIFIED ECURITY CLASSIFICATION OF THIS PAGE IForm Approved REPORT DOCUMENTATION PAGE MB No, 0704-0188 la .

High temperature wear performance of HVOF-sprayed Cr3C2-WC-NiCoCrMo and Cr3C2-NiCr hardmetal coatings

NASA Astrophysics Data System (ADS)

Zhou, Wuxi; Zhou, Kesong; Li, Yuxi; Deng, Chunming; Zeng, Keli

2017-09-01

A novel Cr3C2-WC-NiCoCrMo and commercial Cr3C2-NiCr thermal spray-grade powders with particle size of -45 + 15 μm were prepared by an agglomeration and sintering process. Cr3C2-WC-NiCoCrMo and Cr3C2-NiCr coatings were deposited by high velocity oxygen fuel (HVOF) spraying. The fundamental properties of both coatings were evaluated and friction wear test against Al2O3 counterbodies of both coatings at high temperatures (450 °C, 550 °C, 650 °C) were carried out ball-on-disk high temperature tribometer. All specimens were characterized by optical microscopy, X-ray diffraction (XRD), X-ray photoelectron spectroscopy (XPS), scanning electron microscopy with energy dispersive spectroscopy (SEM/EDS) and 3D non-contact surface mapping profiler. The results have shown that the Cr3C2-WC-NiCoCrMo coating exhibited lower porosity, higher micro-hardness compared to the Cr3C2-NiCr coating. The Cr3C2-WC-NiCoCrMo coating also exhibited better wear resistance and higher friction coefficient compared to the Cr3C2-NiCr coating when sliding against the Al2O3 counterpart. Wear rates of both coatings increased with raising temperature. Both coatings experienced abrasive wear; hard phase particles (WC and Cr3C2) with different sizes, distributed in the matrix phase, will effectively improve the resistance against wear at high temperatures.

Sources and distribution of trace elements in Estonian peat

NASA Astrophysics Data System (ADS)

Orru, Hans; Orru, Mall

2006-10-01

This paper presents the results of the distribution of trace elements in Estonian mires. Sixty four mires, representative of the different landscape units, were analyzed for the content of 16 trace elements (Cr, Mn, Ni, Cu, Zn, and Pb using AAS; Cd by GF-AAS; Hg by the cold vapour method; and V, Co, As, Sr, Mo, Th, and U by XRF) as well as other peat characteristics (peat type, degree of humification, pH and ash content). The results of the research show that concentrations of trace elements in peat are generally low: V 3.8 ± 0.6, Cr 3.1 ± 0.2, Mn 35.1 ± 2.7, Co 0.50 ± 0.05, Ni 3.7 ± 0.2, Cu 4.4 ± 0.3, Zn 10.0 ± 0.7, As 2.4 ± 0.3, Sr 21.9 ± 0.9, Mo 1.2 ± 0.2, Cd 0.12 ± 0.01, Hg 0.05 ± 0.01, Pb 3.3 ± 0.2, Th 0.47 ± 0.05, U 1.3 ± 0.2 μg g - 1 and S 0.25 ± 0.02%. Statistical analyses on these large database showed that Co has the highest positive correlations with many elements and ash content. As, Ni, Mo, ash content and pH are also significantly correlated. The lowest abundance of most trace elements was recorded in mires fed only by precipitation (ombrotrophic), and the highest in mires fed by groundwater and springs (minerotrophic), which are situated in the flood plains of river valleys. Concentrations usually differ between the superficial, middle and bottom peat layers, but the significance decreases depending on the type of mire in the following order: transitional mires - raised bogs - fens. Differences among mire types are highest for the superficial but not significant for the basal peat layers. The use of peat with high concentrations of trace elements in agriculture, horticulture, as fuel, for water purification etc., may pose a risk for humans: via the food chain, through inhalation, drinking water etc.

High density dispersion fuel

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

Hofman, G.L.

1996-09-01

A fuel development campaign that results in an aluminum plate-type fuel of unlimited LEU burnup capability with an uranium loading of 9 grams per cm{sup 3} of meat should be considered an unqualified success. The current worldwide approved and accepted highest loading is 4.8 g cm{sup {minus}3} with U{sub 3}Si{sub 2} as fuel. High-density uranium compounds offer no real density advantage over U{sub 3}Si{sub 2} and have less desirable fabrication and performance characteristics as well. Of the higher-density compounds, U{sub 3}Si has approximately a 30% higher uranium density but the density of the U{sub 6}X compounds would yield the factormore » 1.5 needed to achieve 9 g cm{sup {minus}3} uranium loading. Unfortunately, irradiation tests proved these peritectic compounds have poor swelling behavior. It is for this reason that the authors are turning to uranium alloys. The reason pure uranium was not seriously considered as a dispersion fuel is mainly due to its high rate of growth and swelling at low temperatures. This problem was solved at least for relatively low burnup application in non-dispersion fuel elements with small additions of Si, Fe, and Al. This so called adjusted uranium has nearly the same density as pure {alpha}-uranium and it seems prudent to reconsider this alloy as a dispersant. Further modifications of uranium metal to achieve higher burnup swelling stability involve stabilization of the cubic {gamma} phase at low temperatures where normally {alpha} phase exists. Several low neutron capture cross section elements such as Zr, Nb, Ti and Mo accomplish this in various degrees. The challenge is to produce a suitable form of fuel powder and develop a plate fabrication procedure, as well as obtain high burnup capability through irradiation testing.« less

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

Sulfidation of Co/Al[sub 2]O[sub 3] and CoMo/Al[sub 2]O[sub 3] catalysts studied by Moessbauer emission spectroscopy

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

Craje, M.W.J.; Kraan, A.M. van der; Beer, V.H.J. de

1993-10-01

The structure of hydrodesulfurization catalysts is relevant to many industries. The sulfidation of uncalcined and calcined alumina-supported cobalt and cobalt-molybdenum catalysts was systematically studied by means of in situ Moessbauer emission spectroscopy (MES) at room temperature. The spectra obtained during the stepwise sulfidation of the uncalcined catalysts clearly resemble those observed for carbon-supported ones. Hence, the interpretation of the spectra of the alumina-supported catalysts is based on the conclusions drawn from the MES studies of the carbon-supported catalysts, which are less complex because Co ions do not diffuse into the support. It is demonstrated that not only in sulfided CoMo/Al[submore » 2]O[sub 3], but also in sulfided Co/Al[sub 2]O[sub 3], catalysts Co-sulfide species with a [open quotes]Co-Mo-S[close quotes]-type quadrupole splitting can be formed. It is concluded that the Co-sulfide species formed in sulfided Co/Al[sub 2]O[sub 3] and CoMo/Al[sub 2]O[sub 3] catalysts are essentially the same, only the particle size and ordering of the Co-sulfide species may differ, as in the case of Co/C and CoMo/C catalysts. The function of the Mo, which is present as MoS[sub 2], is merely to stabilize very small Co-sulfide particles, which in the limit contain only one single Co atom. Furthermore, it turns out that the value of the electric quadrupole splitting (Q.S. value) of the Co-sulfide phase in the sulfided catalysts depends on the sulfiding temperature and Co content. This observation leads to the conclusion that large Q.S. values point to the presence of very small Co-sulfide entities or particles (the lower limit being [open quotes]particles[close quotes] containing only one Co atom, such as proposed in the [open quotes]Co-Mo-S[close quotes] model), whereas small Q.S. values point to the presence of large Co-sulfide particles (the upper limit being crystalline Co[sub 9]S[sub 8]). 28 refs., 7 figs., 6 tabs.« less

Issues and Potential Program on Denatured Fuel Utilization.

DTIC Science & Technology

1978-12-01

HTGR fuel develop - ment program ; 4. coated particles of (U,Th)02 have been extensively tested as potential HTGR fuels . A detailed summary of the...current scrap and waste treatment requirements. dBase case for all HTGR (Prismatic Fuel Element) cases based on data in "Summary Program Plan...Alternate Program for HTGR Fuel Recycle," April 11, 1975, Draft. 19 a --- AC8NCi09 The principal factors that result in a nominally-higher cost for

Application of the DART Code for the Assessment of Advanced Fuel Behavior

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

Rest, J.; Totev, T.

2007-07-01

The Dispersion Analysis Research Tool (DART) code is a dispersion fuel analysis code that contains mechanistically-based fuel and reaction-product swelling models, a one dimensional heat transfer analysis, and mechanical deformation models. DART has been used to simulate the irradiation behavior of uranium oxide, uranium silicide, and uranium molybdenum aluminum dispersion fuels, as well as their monolithic counterparts. The thermal-mechanical DART code has been validated against RERTR tests performed in the ATR for irradiation data on interaction thickness, fuel, matrix, and reaction product volume fractions, and plate thickness changes. The DART fission gas behavior model has been validated against UO{sub 2}more » fission gas release data as well as measured fission gas-bubble size distributions. Here DART is utilized to analyze various aspects of the observed bubble growth in U-Mo/Al interaction product. (authors)« less

Kinetic precipitation of solution-phase polyoxomolybdate followed by transmission electron microscopy: a window to solution-phase nanostructure.

PubMed

Zhu, Yan; Cammers-Goodwin, Arthur; Zhao, Bin; Dozier, Alan; Dickey, Elizabeth C

2004-05-17

This study aimed to elucidate the structural nature of the polydisperse, nanoscopic components in the solution and the solid states of partially reduced polyoxomolybdate derived from the [Mo132] keplerate, [(Mo)Mo5]12-[Mo2 acetate]30. Designer tripodal hexamine-tris-crown ethers and nanoscopic molybdate coprecipitated from aqueous solution. These microcrystalline solids distributed particle radii between 2-30 nm as assayed by transmission electron microscopy (TEM). The solid materials and their particle size distributions were snap shots of the solution phase. The mother liquor of the preparation of the [Mo132] keplerate after three days revealed large species (r=20-30 nm) in the coprecipitate, whereas [Mo132] keplerate redissolved in water revealed small species (3-7 nm) in the coprecipitate. Nanoparticles of coprecipitate were more stable than solids derived solely from partially reduced molybdate. The TEM features of all material analyzed lacked facets on the nanometer length scale; however, the structures diffracted electrons and appeared to be defect-free as evidenced by Moiré patterns in the TEM images. Moiré patterns and size-invariant optical densities of the features in the micrographs suggested that the molybdate nanoparticles were vesicular.

Multi-Megawatt Space Nuclear Power Generation

DTIC Science & Technology

1993-06-28

electric generation, both for open- and closed-cycle opera- tion. These reactors use the particulate fuel of the type developed for HTGR reactors. What...commercial HTGR power reactors, the particles are held in place and directly cooled. Figure 2.7 shows the two types of fuel particles developed for...of MW(e), for pulsed energy devices. The FBR would use HTGR -type particle fuel , contained in a annular bed be- tween two porous frits. Helium would

Two-Dimensional Mapping of the Calculated Fission Power for the Full-Size Fuel Plate Experiment Irradiated in the Advanced Test Reactor

NASA Astrophysics Data System (ADS)

Chang, G. S.; Lillo, M. A.

2009-08-01

The National Nuclear Security Administrations (NNSA) Reduced Enrichment for Research and Test Reactors (RERTR) program assigned to the Idaho National Laboratory (INL) the responsibility of developing and demonstrating high uranium density research reactor fuel forms to enable the use of low enriched uranium (LEU) in research and test reactors around the world. A series of full-size fuel plate experiments have been proposed for irradiation testing in the center flux trap (CFT) position of the Advanced Test Reactor (ATR). These full-size fuel plate tests are designated as the AFIP tests. The AFIP nominal fuel zone is rectangular in shape having a designed length of 21.5-in (54.61-cm), width of 1.6-in (4.064-cm), and uniform thickness of 0.014-in (0.03556-cm). This gives a nominal fuel zone volume of 0.482 in3 (7.89 cm3) per fuel plate. The AFIP test assembly has two test positions. Each test position is designed to hold 2 full-size plates, for a total of 4 full-size plates per test assembly. The AFIP test plates will be irradiated at a peak surface heat flux of about 350 W/cm2 and discharged at a peak U-235 burn-up of about 70 at.%. Based on limited irradiation testing of the monolithic (U-10Mo) fuel form, it is desirable to keep the peak fuel temperature below 250°C to achieve this, it will be necessary to keep plate heat fluxes below 500 W/cm2. Due to the heavy U-235 loading and a plate width of 1.6-in (4.064-cm), the neutron self-shielding will increase the local-to-average-ratio (L2AR) fission power near the sides of the fuel plates. To demonstrate that the AFIP experiment will meet the ATR safety requirements, a very detailed 2-dimensional (2D) Y-Z fission power profile was evaluated in order to best predict the fuel plate temperature distribution. The ability to accurately predict fuel plate power and burnup are essential to both the design of the AFIP tests as well as evaluation of the irradiated fuel performance. To support this need, a detailed MCNP Y-Z mini-plate fuel model was developed. The Y-Z model divides each fuel plate into 30 equal intervals in both the Y and Z directions. The MCNP-calculated results and the detailed Y-Z fission power mapping were used to help design the AFIP fuel test assembly to demonstrate that the AFIP test assembly thermal-hydraulic limits will not exceed the ATR safety limits.

Preparation of U.sub.3 O.sub.8

DOEpatents

Johnson, David R.

1980-01-01

A method is described for the preparation of U.sub.3 O.sub.8 nuclear fuel material by direct precipitation of uranyl formate monohydrate from uranyl nitrate solution. The uranyl formate monohydrate precipitate is removed, dried and calcined to produce U.sub.3 O.sub.8 having a controlled particle size distribution.

AGR-2 Irradiation Test Final As-Run Report, Rev 2

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

Collin, Blaise P.

2014-08-01

This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO 2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. (b) Provide irradiated fuel samplesmore » for post-irradiation experiment (PIE) and safety testing. (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO 2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO 2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel. In order to achieve the test objectives, the AGR-2 experiment was irradiated in the B-12 position of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for a total irradiation duration of 559.2 effective full power days (EFPD). Irradiation began on June 22, 2010, and ended on October 16, 2013, spanning 12 ATR power cycles and approximately three and a half calendar years. The test contained six independently controlled and monitored capsules. Each U.S. capsule contained 12 compacts of either UCO or UO2 AGR coated fuel. No fuel particles failed during the AGR-2 irradiation. Final burnup values on a per compact basis ranged from 7.26 to 13.15% FIMA (fissions per initial heavy-metal atom) for UCO fuel, and 9.01 to 10.69% FIMA for UO 2 fuel, while fast fluence values ranged from 1.94 to 3.47 x 10 25 n/m 2 (E >0.18 MeV) for UCO fuel, and from 3.05 to 3.53 x 10 25 n/m 2 (E >0.18 MeV) for UO 2 fuel. Time-average volume-average (TAVA) temperatures on a capsule basis at the end of irradiation ranged from 987°C in Capsule 6 to 1296°C in Capsule 2 for UCO, and from 996 to 1062°C in UO 2-fueled Capsule 3. By the end of the irradiation, all of the installed thermocouples (TCs) had failed. Fission product release-to-birth (R/B) ratios were quite low. In the UCO capsules, R/B values during the first three cycles were below 10 -6 with the exception of the hotter Capsule 2, in which the R/Bs reached 2 x 10 -6. In the UO 2 capsule (Capsule 3), the R/B values during the first three cycles were below 10 -7. R/B values for all following cycles are not reliable due to gas flow and cross talk issues.« less

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

PubMed

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

2012-01-01

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

Facile synthesis of high surface area molybdenum nitride and carbide

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

Roy, Aaron; Serov, Alexey; Artyushkova, Kateryna

2015-08-15

The synthesis of high surface area γ-Mo{sub 2}N and α-Mo{sub 2}C is reported (116 and 120 m{sup 2}/g) without the temperature programmed reduction of MoO{sub 3}. γ-Mo{sub 2}N was prepared in an NH{sub 3}-free synthesis using forming gas (7 at% H{sub 2}, N{sub 2}-balance) as the reactive atmosphere. Three precursors were studied ((NH{sub 4}){sub 6}Mo{sub 7}O{sub 24}·4H{sub 2}O, (NH{sub 4}){sub 2} Mg(MoO{sub 4}){sub 2}, and MgMoO{sub 4}) along with the sacrificial support method (SSM) as a means of reducing the particle size of Mo{sub 2}N and Mo{sub 2}C. In situ X-ray diffraction (XRD) studies were carried out to identify reactionmore » intermediates, the temperature at which various intermediates form, and the average domain size of the Mo{sub 2}N products. Materials were synthesized in bulk and further characterized by XRD, HRTEM, XPS, and BET. - Highlights: • Facile synthesis of γ-Mo2N and α-Mo2C with surface area exceeding 100 m{sup 2}/g. • Sacrificial support method was used to achieve these high surface areas. • Materials can serve as catalysts or supports in (electro)chemical processes.« less

Utilization of thorium and U-ZrH1.6 fuels in various heterogeneous cores for TRIGA PUSPATI Reactor (RTP)

NASA Astrophysics Data System (ADS)

Damahuri, Abdul Hannan Bin; Mohamed, Hassan; Aziz Mohamed, Abdul; Idris, Faridah

2018-01-01

The use of thorium as nuclear fuel has been an appealing prospect for many years and will be great significance to nuclear power generation. There is an increasing need for more research on thorium as Malaysian government is currently active in the national Thorium Flagship Project, which was launched in 2014. The thorium project, which is still in phase 1, focuses on the research and development of the thorium extraction from mineral processing ore. Thus, the aim of the study is to investigate other alternative TRIGA PUSPATI Reactor (RTP) core designs that can fully utilize thorium. Currently, the RTP reactor has an average neutron flux of 2.797 x 1012 cm-2/s-1 and an effective multiplication factor, k eff, of 1.001. The RTP core has a circular array core configuration with six circular rings. Each ring consists of 6, 12, 18, 24, 30 or 36 U-ZrH1.6 fuel rods. There are three main type of uranium weight, namely 8.5, 12 and 20 wt.%. For this research, uranium zirconium hydride (U-ZrH1.6) fuel rods in the RTP core were replaced by thorium (ThO2) fuel rods. Seven core configurations with different thorium fuel rods placements were modelled in a 2D structure and simulated using Monte Carlo n-particle (MCNPX) code. Results show that the highest initial criticality obtained is around 1.35101. Additionally there is a significant discrepancy between results from previous study and the work because of the large estimated leakage probability of approximately 21.7% and 2D model simplification.

Lactate production by swine adipocytes: effects of age, nutritional status, glucose concentration, and insulin.

PubMed

Heckler, B K; Carey, G B

1997-06-01

To develop an alternative model in which to study the relationship between adipose tissue lactate production, obesity, and non-insulin-dependent diabetes mellitus (NIDDM), we investigated lactate production by swine adipocytes. Subcutaneous adipocytes from fasted 3-wk-old, fasted 7-mo-old, and fed 7-mo-old Yucatan minIature swine were isolated and incubated with 0.2, 1, 5, 10, or 25 mM glucose +/- 1 mU/ml insulin. Total glucose metabolism (TGM) was estimated by product summation. Results showed that 1) TGM was threefold greater in cells from fasted 7-mo- vs. 3-wk-old swine (P < 0.05), 2) TGM was 2.7-fold greater in cells from fed 7-mo-old vs. fasted 7-mo-old swine (P < 0.05), 3) insulin failed to stimulate TGM in adipocytes from swine of either age and either nutritional status, and 4) lactate and pyruvate accounted for 34 and 30% of TGM, respectively, in adipocytes from swine of both ages. Similarities in glucose metabolism and lactate production in adipocytes from swine and obese NIDDM humans make the swine a potentially valuable model for studying lactate production associated with obesity and NIDDM.

High-performance transition metal-doped Pt 3Ni octahedra for oxygen reduction reaction

DOE PAGES

Huang, Xiaoqing; Zhao, Zipeng; Cao, Liang; ...

2015-06-11

Bimetallic platinum-nickel (Pt-Ni) nanostructures represent an emerging class of electrocatalysts for oxygen reduction reaction (ORR) in fuel cells, but practical applications have been limited by catalytic activity and durability. We surface-doped Pt 3Ni octahedra supported on carbon with transition metals, termed M-Pt 3Ni/C, where M is vanadium, chromium, manganese, iron, cobalt, molybdenum (Mo), tungsten, or rhenium. The Mo-Pt 3Ni/C showed the best ORR performance, with a specific activity of 10.3 mA/cm2 and mass activity of 6.98 A/mgPt, which are 81- and 73-fold enhancements compared with the commercial Pt/C catalyst (0.127 mA/cm 2 and 0.096 A/mg Pt). In conclusion, theoretical calculationsmore » suggest that Mo prefers subsurface positions near the particle edges in vacuum and surface vertex/edge sites in oxidizing conditions, where it enhances both the performance and the stability of the Pt3Ni catalyst.« less

Next generation fuel irradiation capability in the High Flux Reactor Petten

NASA Astrophysics Data System (ADS)

Fütterer, Michael A.; D'Agata, Elio; Laurie, Mathias; Marmier, Alain; Scaffidi-Argentina, Francesco; Raison, Philippe; Bakker, Klaas; de Groot, Sander; Klaassen, Frodo

2009-07-01

This paper describes selected equipment and expertise on fuel irradiation testing at the High Flux Reactor (HFR) in Petten, The Netherlands. The reactor went critical in 1961 and holds an operating license up to at least 2015. While HFR has initially focused on Light Water Reactor fuel and materials, it also played a decisive role since the 1970s in the German High Temperature Reactor (HTR) development program. A variety of tests related to fast reactor development in Europe were carried out for next generation fuel and materials, in particular for Very High Temperature Reactor (V/HTR) fuel, fuel for closed fuel cycles (U-Pu and Th-U fuel cycle) and transmutation, as well as for other innovative fuel types. The HFR constitutes a significant European infrastructure tool for the development of next generation reactors. Experimental facilities addressed include V/HTR fuel tests, a coated particle irradiation rig, and tests on fast reactor, transmutation and thorium fuel. The rationales for these tests are given, results are provided and further work is outlined.

Fabrication of Monolithic RERTR Fuels by Hot Isostatic Pressing

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

Jan-Fong Jue; Blair H. Park; Curtis R. Clark

2010-11-01

The RERTR (Reduced Enrichment for Research and Test Reactors) Program is developing advanced nuclear fuels for high-power test reactors. Monolithic fuel design provides higher uranium loading than that of the traditional dispersion fuel design. Hot isostatic pressing is a promising process for low-cost batch fabrication of monolithic RERTR fuel plates for these high-power reactors. Bonding U Mo fuel foil and 6061 Al cladding by hot isostatic press bonding was successfully developed at Idaho National Laboratory. Due to the relatively high processing temperature, the interaction between fuel meat and aluminum cladding is a concern. Two different methods were employed to mitigatemore » this effect: (1) a diffusion barrier and (2) a doping addition to the interface. Both types of fuel plates have been fabricated by hot isostatic press bonding. Preliminary results show that the direct fuel/cladding interaction during the bonding process was eliminated by introducing a thin zirconium diffusion barrier layer between the fuel and the cladding. Fuel plates were also produced and characterized with a silicon-rich interlayer between fuel and cladding. This paper reports the recent progress of this developmental effort and identifies the areas that need further attention.« less

Environmental toxicology and risk assessment: Seventh volume

USGS Publications Warehouse

Little, Edward E.; Greenberg, Bruce M.; DeLonay, Aaron J.

1998-01-01

This publication, Environmental Toxicology and Risk Assessment: Seventh Volume, contains papers presented at the Seventh Symposium on Toxicology and Risk Assessment: Ultraviolet Radiation and the Environment, held 7-9 April, 1997 in St. Louis, MO. The symposium, the 24th in a series on environmental toxicology, was sponsored by Committee E-47. Edward E. Little, of the U.S. Geological Survey/Biological Services Division in Columbia, MO, presided as chairman of the symposium. Bruce M. Greenberg, with the Department of Biology at the University of Waterloo in Ontario, Canada, and Aaron J. DeLonay, also with the U.S. Geological Service/Biological Services Division in Columbia, MO, served as co-chairmen of the symposium. Each of these men served as editor of the resulting publication.

Structure and electronic properties of Cu nanoclusters supported on Mo 2C(001) and MoC(001) surfaces

DOE PAGES

Posada-Pérez, Sergio; Viñes, Francesc; Rodríguez, José A.; ...

2015-09-15

In this study, the atomic structure and electronic properties of Cu n nanoclusters (n = 4, 6, 7, and 10) supported on cubic nonpolar δ-MoC(001) and orthorhombic C- or Mo-terminated polar β-Mo 2C(001) surfaces have been investigated by means of periodic density functional theory based calculations. The electronic properties have been analyzed by means of the density of states, Bader charges, and electron localization function plots. The Cu nanoparticles supported on β-Mo 2C(001), either Mo- or C-terminated, tend to present a two-dimensional structure whereas a three-dimensional geometry is preferred when supported on δ-MoC(001), indicating that the Mo:C ratio and themore » surface polarity play a key role determining the structure of supported clusters. Nevertheless, calculations also reveal important differences between the C- and Mo-terminated β-Mo 2C(001) supports to the point that supported Cu particles exhibit different charge states, which opens a way to control the reactivity of these potential catalysts.« less

Use of Moringa oleifera seed extracts to reduce helminth egg numbers and turbidity in irrigation water.

PubMed

Sengupta, Mita E; Keraita, Bernard; Olsen, Annette; Boateng, Osei K; Thamsborg, Stig M; Pálsdóttir, Guðný R; Dalsgaard, Anders

2012-07-01

Water from wastewater-polluted streams and dug-outs is the most commonly used water source for irrigation in urban farming in Ghana, but helminth parasite eggs in the water represent health risks when used for crop production. Conventional water treatment is expensive, requires advanced technology and often breaks down in less developed countries so low cost interventions are needed. Field and laboratory based trials were carried out in order to investigate the effect of the natural coagulant Moringa oleifera (MO) seed extracts in reducing helminh eggs and turbidity in irrigation water, turbid water, wastewater and tap water. In medium to high turbid water MO extracts were effective in reducing the number of helminth eggs by 94-99.5% to 1-2 eggs per litre and the turbidity to 7-11 NTU which is an 85-96% reduction. MO is readily available in many tropical countries and can be used by farmers to treat high turbid water for irrigation, however, additional improvements of water quality, e.g. by sand filtration, is suggested to meet the guideline value of ≤ 1 helminth egg per litre and a turbidity of ≤ 2 NTU as recommended by the World Health Organization and the U.S. Environmental Protection Agency for water intended for irrigation. A positive correlation was established between reduction in turbidity and helminth eggs in irrigation water, turbid water and wastewater treated with MO. This indicates that helminth eggs attach to suspended particles and/or flocs facilitated by MO in the water, and that turbidity and helminth eggs are reduced with the settling flocs. However, more experiments with water samples containing naturally occurring helminth eggs are needed to establish whether turbidity can be used as a proxy for helminth eggs. Copyright © 2012 Elsevier Ltd. All rights reserved.

LANL Experience Rolling Zr-Clad LEU-10Mo Foils for AFIP-7

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

Hammon, Duncan L.; Clarke, Kester D.; Alexander, David J.

2015-05-29

The cleaning, canning, rolling and final trimming of Low Enriched Uranium-10 wt. pct. Molybdenum (LEU-10Mo) foils for ATR (Advanced Test Reactor) fuel plates to be used in the AFIP-7 (ATR Full Size Plate In Center Flux Trap Position) experiments are summarized. Six Zr-clad foils were produced from two LEU-10Mo castings supplied to Los Alamos National Laboratory (LANL) by Y-12 National Security Complex. Details of cleaning and canning procedures are provided. Hot- and cold-rolling results are presented, including rolling schedules, images of foils in-process, metallography and local compositions of regions of interest, and details of final foil dimensions and process yield.more » This report was compiled from the slides for the presentation of the same name given by Duncan Hammon on May 12, 2011 at the AFIP-7 Lessons Learned meeting in Salt Lake City, UT, with Los Alamos National Laboratory document number LA-UR 11-02898.« less

Method for the production of .sup.99m Tc compositions from .sup.99 Mo-containing materials

DOEpatents

Bennett, Ralph G.; Christian, Jerry D.; Grover, S. Blaine; Petti, David A.; Terry, William K.; Yoon, Woo Y.

1998-01-01

An improved method for producing .sup.99m Tc compositions from .sup.99 Mo compounds. .sup.100 Mo metal or .sup.100 MoO.sub.3 is irradiated with photons in a particle (electron) accelerator to ultimately produce .sup.99 MoO.sub.3. This composition is then heated in a reaction chamber to form a pool of molten .sup.99 MoO.sub.3 with an optimum depth of 0.5-5 mm. A gaseous mixture thereafter evolves from the molten .sup.99 MoO.sub.3 which contains vaporized .sup.99 MoO.sub.3, vaporized .sup.99m TcO.sub.3, and vaporized .sup.99m TcO.sub.2. This mixture is then combined with an oxidizing gas (O.sub.2(g)) to generate a gaseous stream containing vaporized .sup.99m Tc.sub.2 O.sub.7 and vaporized .sup.99 MoO.sub.3. Next, the gaseous stream is cooled in a primary condensation stage in the reaction chamber to remove vaporized .sup.99 MoO.sub.3. Cooling is undertaken at a specially-controlled rate to achieve maximum separation efficiency. The gaseous stream is then cooled in a sequential secondary condensation stage to convert vaporized .sup.99m Tc.sub.2 O.sub.7 into a condensed .sup.99m Tc-containing reaction product which is collected.

Method for the production of {sup 99m}Tc compositions from {sup 99}Mo-containing materials

DOEpatents

Bennett, R.G.; Christian, J.D.; Grover, S.B.; Petti, D.A.; Terry, W.K.; Yoon, W.Y.

1998-09-01

An improved method is described for producing {sup 99m}Tc compositions from {sup 99}Mo compounds. {sup 100}Mo metal or {sup 100}MoO{sub 3} is irradiated with photons in a particle (electron) accelerator to ultimately produce {sup 99}MoO{sub 3}. This composition is then heated in a reaction chamber to form a pool of molten {sup 99}MoO{sub 3} with an optimum depth of 0.5--5 mm. A gaseous mixture thereafter evolves from the molten {sup 99}MoO{sub 3} which contains vaporized {sup 99}MoO{sub 3}, vaporized {sup 99m}TcO{sub 3}, and vaporized {sup 99m}TcO{sub 2}. This mixture is then combined with an oxidizing gas (O{sub 2(g)}) to generate a gaseous stream containing vaporized {sup 99m}Tc{sub 2}O{sub 7} and vaporized {sup 99}MoO{sub 3}. Next, the gaseous stream is cooled in a primary condensation stage in the reaction chamber to remove vaporized {sup 99}MoO{sub 3}. Cooling is undertaken at a specially-controlled rate to achieve maximum separation efficiency. The gaseous stream is then cooled in a sequential secondary condensation stage to convert vaporized {sup 99m}Tc{sub 2}O{sub 7} into a condensed {sup 99m}Tc-containing reaction product which is collected. 1 fig.

Effects of the foil flatness on the stress-strain characteristics of U10Mo alloy based monolithic mini-plates

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

Hakan Ozaltun; Pavel Medvedev

The effects of the foil flatness on stress-strain behavior of monolithic fuel mini-plates during fabrication and irradiation were studied. Monolithic plate-type fuels are a new fuel form being developed for research and test reactors to achieve higher uranium densities. This concept facilitates the use of low-enriched uranium fuel in the reactor. These fuel elements are comprised of a high density, low enrichment, U–Mo alloy based fuel foil encapsulated in a cladding material made of Aluminum. To evaluate the effects of the foil flatness on the stress-strain behavior of the plates during fabrication, irradiation and shutdown stages, a representative plate frommore » RERTR-12 experiments (Plate L1P756) was considered. Both fabrication and irradiation processes of the plate were simulated by using actual irradiation parameters. The simulations were repeated for various foil curvatures to observe the effects of the foil flatness on the peak stress and strain magnitudes of the fuel elements. Results of fabrication simulations revealed that the flatness of the foil does not have a considerable impact on the post fabrication stress-strain fields. Furthermore, the irradiation simulations indicated that any post-fabrication stresses in the foil would be relieved relatively fast in the reactor. While, the perfectly flat foil provided the slightly better mechanical performance, overall difference between the flat-foil case and curved-foil case was not significant. Even though the peak stresses are less affected, the foil curvature has several implications on the strain magnitudes in the cladding. It was observed that with an increasing foil curvature, there is a slight increase in the cladding strains.« less

Controlled synthesis and luminescence properties of CaMoO4:Eu3+ microcrystals

NASA Astrophysics Data System (ADS)

Xie, Ying; Ma, Siming; Wang, Yu; Xu, Mai; Lu, Chengxi; Xiao, Linjiu; Deng, Shuguang

2018-03-01

Pure tetragonal-phased Ca0.9MoO4:0.1Eu3+ (CaMoO4:Eu3+) microcrystals with varying particle sizes were prepared via a co-deposition in water/oil (w/o) phase method. The particle sizes of as-prepared samples were controlled by calcination temperature and calcination time, and the crystallinity of the samples enhances with increasing particle size. The luminescence properties of CaMoO4:Eu3+ microcrystals were studied with varying particle size. The results reveal that the intensity of emission spectra of the CaMoO4:Eu3+ samples increases with increasing particle size, and they have closely correlation with each other. It is the same with the luminescence lifetime. The luminescence lifetime of the CaMoO4:Eu3+ samples decreases from 0.637 ms to 0.447 ms with increasing particle size from 0.12 μm to 1.79 μm, respectively. This study not only provides information for size-dependent luminescence properties of CaMoO4:Eu3+ but also gives a reference for potential applications in high voltage electric porcelain material.

Method for generating a crystalline .sup.99 MoO.sub.3 product and the isolation .sup.99m Tc compositions therefrom

DOEpatents

Bennett, Ralph G.; Christian, Jerry D.; Kirkham, Robert J.; Tranter, Troy J.

1998-01-01

An improved method for producing .sup.99m Tc compositions. .sup.100 Mo metal is irradiated with photons in a particle (electron) accelerator to produce .sup.99 Mo metal which is dissolved in a solvent. A solvated .sup.99 Mo product is then dried to generate a supply of .sup.99 MoO.sub.3 crystals. The crystals are thereafter heated at a temperature which will sublimate the crystals and form a gaseous mixture containing vaporized .sup.99m TcO.sub.3 and vaporized .sup.99m TcO.sub.2 but will not cause the production of vaporized .sup.99 MoO.sub.3. The mixture is then combined with an oxidizing gas to generate a gaseous stream containing vaporized .sup.99m Tc.sub.2 O.sub.7. Next, the gaseous stream is cooled to a temperature sufficient to convert the vaporized .sup.99m Tc.sub.2 O.sub.7 into a condensed .sup.99m Tc-containing product. The product has high purity levels resulting from the use of reduced temperature conditions and ultrafine crystalline .sup.99 MoO.sub.3 starting materials with segregated .sup.99m Tc compositions therein which avoid the production of vaporized .sup.99 MoO.sub.3 contaminants.

RERTR-6 Irradiation Summary Report

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

D. M. Perez; M. A. Lillo; G. S. Chang

2011-12-01

The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-6 was designed to evaluate several modified fuel designs that were proposed to address the possibility of breakaway swelling due to porosity within the (U. Mo) Al interaction product observed in the full-size plate tests performed in Russia and France1. The following report summarizes the life of the RERTR-6 experiment through end of irradiation, including as-run neutronic analyses, thermal analyses and hydraulic testing results.

Generation and characterization of diesel engine combustion emissions from petroleum diesel and soybean biodiesel fuels and application for inhalation exposure studies.

PubMed

Mutlu, Esra; Nash, David G; King, Charly; Krantz, Todd Q; Preston, William T; Kooter, Ingeborg M; Higuchi, Mark; DeMarini, David; Linak, William P; Gilmour, M Ian

2015-01-01

Biodiesel made from the transesterification of plant- and animal-derived oils is an important alternative fuel source for diesel engines. Although numerous studies have reported health effects associated with petroleum diesel emissions, information on biodiesel emissions are more limited. To this end, a program at the U.S. EPA assessed health effects of biodiesel emissions in rodent inhalation models. Commercially obtained soybean biodiesel (B100) and a 20% blend with petroleum diesel (B20) were compared to pure petroleum diesel (B0). Rats and mice were exposed independently for 4 h/day, 5 days/week for up to 6 weeks. Exposures were controlled by dilution air to obtain low (50 µg/m(3)), medium (150 µg/m(3)) and high (500 µg/m(3)) diesel particulate mass (PM) concentrations, and compared to filtered air. This article provides details on facilities, fuels, operating conditions, emission factors and physico-chemical characteristics of the emissions used for inhalation exposures and in vitro studies. Initial engine exhaust PM concentrations for the B100 fuel (19.7 ± 0.7 mg/m(3)) were 30% lower than those of the B0 fuel (28.0 ± 1.5 mg/m(3)). When emissions were diluted with air to control equivalent PM mass concentrations, B0 exposures had higher CO and slightly lower NO concentrations than B100. Organic/elemental carbon ratios and oxygenated methyl esters and organic acids were higher for the B100 than B0. Both the B0 and B100 fuels produced unimodal-accumulation mode particle-size distributions, with B0 producing lower concentrations of slightly larger particles. Subsequent papers in this series will describe the effects of these atmospheres on cardiopulmonary responses and in vitro genotoxicity studies.

Activity and Stability of Dispersed Multi Metallic Pt-based Catalysts for CO Tolerance in Proton Exchange Membrane Fuel Cell Anodes.

PubMed

Hassan, Ayaz; Ticianelli, Edson A

2018-01-01

Studies aiming at improving the activity and stability of dispersed W and Mo containing Pt catalysts for the CO tolerance in proton exchange membrane fuel cell (PEMFC) anodes are revised for the following catalyst systems: (1) a carbon supported PtMo electrocatalyst submitted to heat treatments; (2) Pt and PtMo nanoparticles deposited on carbon-supported molybdenum carbides (Mo2C/C); (3) ternary and quaternary materials formed by PtMoFe/C, PtMoRu/C and PtMoRuFe/C and; (4) Pt nanoparticles supported on tungsten carbide/carbon catalysts and its parallel evaluation with carbon supported PtW catalyst. The heat-treated (600 oC) Pt-Mo/C catalyst showed higher hydrogen oxidation activity in the absence and in the presence of CO and better stability, compared to all other Mo-containing catalysts. PtMoRuFe, PtMoFe, PtMoRu supported on carbon and Pt supported on Mo2C/C exhibited similar CO tolerances but better stability, as compared to as-prepared PtMo supported on carbon. Among the tungsten-based catalysts, tungsten carbide supported Pt catalyst showed reasonable performance and reliable stability in comparison to simple carbon supported PtW catalyst, though an uneven level of catalytic activity towards H2 oxidation in presence of CO is observed for the former as compared to Mo containing catalyst. However, a small dissolution of Mo, Ru, Fe and W from the anodes and their migration toward cathodes during the cell operation is observed. These results indicate that the fuel cell performance and stability has been improved but not yet totally resolved.

Probing Interaction Between Platinum Group Metal (PGM) and Non-PGM Support Through Surface Characterization and Device Performance

NASA Astrophysics Data System (ADS)

Saha, Shibely

High cost and limited abundance of Platinum (Pt) have hindered effective commercialization of Proton Exchange Membrane Fuel Cell and Electrolyzer. Efforts have been undertaken to reduce precious group metal (PGM) requirement for these devices without compromising the activity of the catalyst by using transition metal carbides (TMC) as non-PGM support thanks to their similar electronic and geometric structures as Pt. In this work Mo2C was selected as non-PGM support and Pt was used as the PGM of interest. We hypothesize that the hollow nanotube morphology of Mo2C support combined with Pt nano particles deposited on it via atomic layer deposition (ALD) technique would allow increased interaction between them which may increase the activity of Pt and Mo2C as well as maximize the Pt active surface area. Specifically, a rotary ALD equipment was used to grow Pt particles from atomic level to 2--3 nanometers by simply adjusting number of ALD cycles in order to probe the interaction between the deposited Pt nanoparticles and Mo2C nanotube support. Interaction between the Pt and Mo2 C was analyzed via surface characterization and electrochemical characterization. Interaction between Pt and Mo2C arises due to the lattice mismatch between Pt and Mo2C as well as electron migration between them. Lattice spacing analysis using high resolution transmission electron microscopy (HRTEM) images, combined with Pt binding energy shift in XPS results, clearly showed strong bonding between Pt nanoparticles and the Mo2C nanotube support in all the resultant Pt/Mo2C samples. We postulate that this strong interaction is responsible for the significantly enhanced durability observed in our constant potential electrolysis (CPE) and accelerated degradation testing (ADT). Of the three samples from different ALD cycles (15, 50 and 100), Mo2C nanotubes modified by 50 (1.07 wt% Pt loading) and 100 cycles (4.4 wt% Pt) of Pt deposition, showed higher HER and HOR activity per Pt mass than commercial 20% Pt supported on carbon black. Finally, we report the systematic investigation of the feasibility of this nanoscale Pt/Mo 2C catalyst in a practical device setting. The ORR activity of 100 Pt/Mo 2C was determined using the catalyst in the cathode of the MEA. Performance of this catalyst led the Pt utilization to be 10.35kWgPt-1 outperforming the target set by DOE for 2017--2020 by 30%.

Direct methanol fuel cell with extended reaction zone anode: PtRu and PtRuMo supported on graphite felt

NASA Astrophysics Data System (ADS)

Bauer, Alex; Gyenge, Előd L.; Oloman, Colin W.

Pressed graphite felt (thickness ∼350 μm) with electrodeposited PtRu (43 g m -2, 1.4:1 atomic ratio) or PtRuMo (52 g m -2, 1:1:0.3 atomic ratio) nanoparticle catalysts was investigated as an anode for direct methanol fuel cells. At temperatures above 333 K the fuel cell performance of the PtRuMo catalyst was superior compared to PtRu. The power density was 2200 W m -2 with PtRuMo at 5500 A m -2 and 353 K while under the same conditions PtRu yielded 1925 W m -2. However, the degradation rate of the Mo containing catalyst formulation was higher. Compared to conventional gas diffusion electrodes with comparable PtRu catalyst composition and load, the graphite felt anodes gave higher power densities mainly due to the extended reaction zone for methanol oxidation.

Hot Isostatic Press Manufacturing Process Development for Fabrication of RERTR Monolithic Fuel Plates

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

Crapps, Justin M.; Clarke, Kester D.; Katz, Joel D.

2012-06-06

We use experimentation and finite element modeling to study a Hot Isostatic Press (HIP) manufacturing process for U-10Mo Monolithic Fuel Plates. Finite element simulations are used to identify the material properties affecting the process and improve the process geometry. Accounting for the high temperature material properties and plasticity is important to obtain qualitative agreement between model and experimental results. The model allows us to improve the process geometry and provide guidance on selection of material and finish conditions for the process strongbacks. We conclude that the HIP can must be fully filled to provide uniform normal stress across the bondingmore » interface.« less

PRIMARY PARTICLES GENERATED BY THE COMBUSTION OF HEAVY FUEL OIL AND COAL: REVIEW OF RESEARCH RESULTS FROM EPA'S NATIONAL RISK MANAGEMENT RESEARCH LABORATORY

EPA Science Inventory

Researchers at the U.S. Environmental Protection Agency's (EPA's) Office of Research and
Development (ORD) have conducted a series of tests to characterize the size and composition of primary particulate matter (PM) generated from the combustion of heavy fuel oil and pulverize...

Mechanical properties of Mo-Si-B alloys fabricated by using core-shell powder with dispersion of yttria nanoparticles

NASA Astrophysics Data System (ADS)

Byun, Jong Min; Bang, Su-Ryong; Choi, Won June; Kim, Min Sang; Noh, Goo Won; Kim, Young Do

2017-01-01

In recent years, refractory materials with excellent high-temperature properties have been in the spotlight as a next generation's high-temperature materials. Among these, Mo-Si-B alloys composed of two intermetallic compound phases (Mo5SiB2 and Mo3Si) and a ductile α-Mo phase have shown an outstanding thermal properties. However, due to the brittleness of the intermetallic compound phases, Mo-Si-B alloys were restricted to apply for the structural materials. So, to enhance the mechanical properties of Mo-Si-B alloys, many efforts to add rare-earth oxide particles in the Mo-Si-B alloy were performed to induce the improvement of strength and fracture toughness. In this study, to investigate the effect of adding nano-sized Y2O3 particles in Mo-Si-B alloy, a core-shell powder consisting of intermetallic compound phases as the core and nano-sized α-Mo and Y2O3 particles surrounding the core was fabricated. Then pressureless sintering was carried out at 1400 °C for 3 h, and the mechanical properties of sintered bodies with different amounts of Y2O3 particles were evaluated by Vickers hardness and 3-point bending test. Vickers hardness was improved by dispersed Y2O3 particles in the Mo-Si-B alloy. Especially, Mo-3Si-1B-1.5Y2O3 alloy had the highest value, 589 Hv. The fracture toughness was measured using Mo-3Si-1B-1.5Y2O3 alloy and the value indicated as 13.5 MPa·√m.

Faster Electron Injection and More Active Sites for Efficient Photocatalytic H2 Evolution in g-C3 N4 /MoS2 Hybrid.

PubMed

Shi, Xiaowei; Fujitsuka, Mamoru; Kim, Sooyeon; Majima, Tetsuro

2018-03-01

Herein, the structural effect of MoS 2 as a cocatalyst of photocatalytic H 2 generation activity of g-C 3 N 4 under visible light irradiation is studied. By using single-particle photoluminescence (PL) and femtosecond time-resolved transient absorption spectroscopies, charge transfer kinetics between g-C 3 N 4 and two kinds of nanostructured MoS 2 (nanodot and monolayer) are systematically investigated. Single-particle PL results show the emission of g-C 3 N 4 is quenched by MoS 2 nanodots more effectively than MoS 2 monolayers. Electron injection rate and efficiency of g-C 3 N 4 /MoS 2 -nanodot hybrid are calculated to be 5.96 × 10 9 s -1 and 73.3%, respectively, from transient absorption spectral measurement, which are 4.8 times faster and 2.0 times higher than those of g-C 3 N 4 /MoS 2 -monolayer hybrid. Stronger intimate junction between MoS 2 nanodots and g-C 3 N 4 is suggested to be responsible for faster and more efficient electron injection. In addition, more unsaturated terminal sulfur atoms can serve as the active site in MoS 2 nanodot compared with MoS 2 monolayer. Therefore, g-C 3 N 4 /MoS 2 nanodot exhibits a 7.9 times higher photocatalytic activity for H 2 evolution (660 µmol g- 1 h -1 ) than g-C 3 N 4 /MoS 2 monolayer (83.8 µmol g -1 h -1 ). This work provides deep insight into charge transfer between g-C 3 N 4 and nanostructured MoS 2 cocatalysts, which can open a new avenue for more rationally designing MoS 2 -based catalysts for H 2 evolution. © 2018 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Fabrication of inorganic molybdenum disulfide fullerenes by arc in water

NASA Astrophysics Data System (ADS)

Sano, Noriaki; Wang, Haolan; Chhowalla, Manish; Alexandrou, Ioannis; Amaratunga, Gehan A. J.; Naito, Masakazu; Kanki, Tatsuo

2003-01-01

Closed caged fullerene-like molybdenum disulfide (MoS 2) nano-particles were obtained via an arc discharge between a graphite cathode and a molybdenum anode filled with microscopic MoS 2 powder submerged in de-ionized water. A statistical study of over 150 polyhedral fullerene-like MoS 2 nano-particles in plan view transmission electron microscopy revealed that the majority consisted of 2-3 layers with diameters of 5-15 nm. We show that the nano-particles are formed by seamless folding of MoS 2 sheets. A model based on the agglomeration of MoS 2 fragments over an extreme temperature gradient around a plasma ball in water is proposed to explain the formation of nano-particles.

From {sub {infinity}}{sup 1}[(UO{sub 2}){sub 2}O(MoO{sub 4}){sub 4}]{sup 6-} to {sub {infinity}}{sup 1}[(UO{sub 2}){sub 2}(MoO{sub 4}){sub 3}(MoO{sub 5})]{sup 6-} infinite chains in A{sub 6}U{sub 2}Mo{sub 4}O{sub 21} (A=Na, K, Rb, Cs) compounds: Synthesis and crystal structure of Cs{sub 6}[(UO{sub 2}){sub 2}(MoO{sub 4}){sub 3}(MoO{sub 5})

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

Yagoubi, S.; Groupe de Radiochimie, Institut de Physique Nucleaire d'Orsay, Universite Paris-Sud XI, 91406 Orsay Cedex; Obbade, S., E-mail: said.obbade@phelma.grenoble-inp.f

2011-05-15

A new caesium uranyl molybdate belonging to the M{sub 6}U{sub 2}Mo{sub 4}O{sub 21} family has been synthesized by solid-state reaction and its structure determined from single-crystal X-ray diffraction data. Contrary to the other alkali uranyl molybdates of this family (A=Na, K, Rb) where molybdenum atoms adopt only tetrahedral coordination and which can be formulated A{sub 6}[(UO{sub 2}){sub 2}O(MoO{sub 4}){sub 4}], the caesium compound Cs{sub 6}U{sub 2}Mo{sub 4}O{sub 21} should be written Cs{sub 6}[(UO{sub 2}){sub 2}(MoO{sub 4}){sub 3}(MoO{sub 5})] with molybdenum atoms in tetrahedral and square pyramidal environments. Cs{sub 6}[(UO{sub 2}){sub 2}(MoO{sub 4}){sub 3}(MoO{sub 5})] crystallizes in the triclinic symmetry withmore » space group P1-bar and a=10.4275(14) A, b=15.075(2) A, c=17.806(2) A, {alpha}=70.72(1){sup o}, {beta}=80.38(1){sup o} and {gamma}=86.39(1){sup o}, V=2604.7(6) A{sup 3}, Z=4, {rho}{sub mes}=5.02(2) g/cm{sup 3} and {rho}{sub cal}=5.08(3) g/cm{sup 3}. A full-matrix least-squares refinement on the basis of F{sup 2} yielded R{sub 1}=0.0464 and wR{sub 2}=0.0950 for 596 parameters with 6964 independent reflections with I{>=}2{sigma}(I) collected on a BRUKER AXS diffractometer with Mo(K{alpha}) radiation and a CCD detector. The crystal structure of Cs compound is characterized by {sub {infinity}}{sup 1}[(UO{sub 2}){sub 2}(MoO{sub 4}){sub 3}(MoO{sub 5})]{sup 6-} parallels chains built from U{sub 2}O{sub 13} dimeric units, MoO{sub 4} tetrahedra and MoO{sub 5} square pyramids, whereas, Na, K and Rb compounds are characterized by {sub {infinity}}{sup 1}[(UO{sub 2}){sub 2}O(MoO{sub 4}){sub 4}]{sup 6-} parallel chains formulated simply of U{sub 2}O{sub 13} units and MoO{sub 4} tetrahedra. Infrared spectroscopy measurements using powdered samples synthesized by solid-state reaction, confirm the structural results. The thermal stability and the electrical conductivity are also studied. The four compounds decompose at low temperature (between 540 and 610 {sup o}C). -- Graphical abstract: The staking of {sub {infinity}}{sup 1}[(UO{sub 2}){sub 2}(MoO{sub 4}){sub 3}(MoO{sub 5})]{sup 6-} infinite uranyl molybdate ribbons in the Cs{sub 6}[(UO{sub 2}){sub 2}(MoO{sub 4}){sub 3}(MoO{sub 5})] structure. Display Omitted Highlights: {yields} Cs{sub 6}U{sub 2}Mo{sub 4}O{sub 2} a new compound with bidimensional crystal structure, characterized by infinite uranyl molybdate chains. {yields} Crystal structure similar to these of the compounds containing Na, K, Rb. {yields} Molybdenum atoms surrounded by five oxygen atoms to form an original and strongly distorted MoO{sub 5} environment. {yields} The chains arrangement illustrates the key role of the alkaline ionic radius, in the crystal structure distortion for Cs compound.« less

Compact toroid injection fueling in a large field-reversed configuration

NASA Astrophysics Data System (ADS)

Asai, T.; Matsumoto, T.; Roche, T.; Allfrey, I.; Gota, H.; Sekiguchi, J.; Edo, T.; Garate, E.; Takahashi, Ts.; Binderbauer, M.; Tajima, T.

2017-07-01

A repetitively driven compact toroid (CT) injector has been developed for the large field-reversed configuration (FRC) facility of the C-2/C-2U, primarily for particle refueling. A CT is formed and injected by a magnetized coaxial plasma gun (MCPG) exclusively developed for the C-2/C-2U FRC. To refuel the particles of long-lived FRCs, multiple CT injections are required. Thus, a multi-stage discharge circuit was developed for a multi-pulsed CT injection. The drive frequency of this system can be adjusted up to 1 kHz and the number of CT shots per injector is two; the system can be further upgraded for a larger number of injection pulses. The developed MCPG can achieve a supersonic ejection velocity in the range of ~100 km s-1. The key plasma parameters of electron density, electron temperature and the number of particles are ~5  ×  1021 m-3, ~30 eV and 0.5-1.0  ×  1019, respectively. In this project, single- and double-pulsed counter CT injection fueling were conducted on the C-2/C-2U facility by two CT injectors. The CT injectors were mounted 1 m apart in the vicinity of the mid-plane. To avoid disruptive perturbation on the FRC, the CT injectors were operated at the lower limit of the particle inventory. The experiments demonstrated successful refueling with a significant density build-up of 20-30% of the FRC particle inventory per single CT injection without any deleterious effects on the C-2/C-2U FRC.

Interim Report on Mixing During the Casting of LEU-10Mo Plates in the Triple Plate Molds

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

Aikin, Jr., Robert M.

LEU-10%Mo castings are commonly produced by down blending unalloyed HEU with a DU-12.7%Mo master-alloy. This work uses process modeling to provide insight into the mixing of the unalloyed uranium and U-Mo master alloy during melting and mold filling of a triple plate casting. Two different sets of situations are considered: (1) mixing during mold filling from a compositionally stratified crucible and (2) convective mixing of a compositionally stratified crucible during mold heating. The mold filling simulations are performed on the original Y-12 triple plate mold and the horizontal triple plate mold.

Use of the Hugoniot elastic limit in laser shockwave experiments to relate velocity measurements

NASA Astrophysics Data System (ADS)

Smith, James A.; Lacy, Jeffrey M.; Lévesque, Daniel; Monchalin, Jean-Pierre; Lord, Martin

2016-02-01

The US National Nuclear Security Agency has a Global Threat Reduction Initiative (GTRI) with the goal of reducing the worldwide use of high-enriched uranium (HEU). A salient component of that initiative is the conversion of research reactors from HEU to low enriched uranium (LEU) fuels. An innovative fuel is being developed to replace HEU in high-power research reactors. The new LEU fuel is a monolithic fuel made from a U-Mo alloy foil encapsulated in Al-6061 cladding. In order to support the fuel qualification process, the Laser Shockwave Technique (LST) is being developed to characterize the clad-clad and fuel-clad interface strengths in fresh and irradiated fuel plates. This fuel-cladding interface qualification will ensure the survivability of the fuel plates in the harsh reactor environment even under abnormal operating conditions. One of the concerns of the project is the difficulty of calibrating and standardizing the laser shock technique. An analytical study under development and experimental testing supports the hypothesis that the Hugoniot Elastic Limit (HEL) in materials can be a robust and simple benchmark to compare stresses generated by different laser shock systems.

Characterization and sintering of MoO{sub 3} prepared by thermal decomposition of hexaammonium heptamolybdate tetrahydrate

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

Honma, K.; Hirota, K.; Yamaguchi, O.

1997-01-01

MoO{sub 3} powders with plate particles have been prepared by heating hexaammonium heptamolybdate tetrahydrate ((NH{sub 4}){sub 6}Mo{sub 7}O{sub 24} {center_dot} 4H{sub 2}O) for 1 h at 650 C. Densification has been performed by hot pressing for 2 h at 630 C and 30 MPa. Grain-oriented MoO{sub 3} ceramics (>98% of theoretical) are fabricated, in which the <010> axis is parallel to the hot pressing direction. Their electrical conductivities have been measured in the temperature range of 100 to 600 C. Sintered MoO{sub 3} is a ceramic semiconductor. Activation energies are determined to be 0.49 and 0.76 eV for initial andmore » final stages, respectively.« 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.

CNEA/ANL collaboration program to develop an optimized version of DART validation and assessment by means of U{sub 3}Si{sub x} and U{sub 3}O{sub 8-}Al dispersed CNEA miniplate irradiation behavior.

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

Solis, D.

1998-10-16

The DART code is based upon a thermomechanical model that can predict swelling, recrystallization, fuel-meat interdiffusion and other issues related with MTR dispersed FE behavior under irradiation. As a part of a common effort to develop an optimized version of DART, a comparison between DART predictions and CNEA miniplates irradiation experimental data was made. The irradiation took place during 1981-82 for U3O8 miniplates and 1985-86 for U{sub 3}Si{sub x} at Oak Ridge Research Reactor (ORR). The microphotographs were studied by means of IMAWIN 3.0 Image Analysis Code and different fission gas bubbles distributions were obtained. Also it was possible tomore » find and identify different morphologic zones. In both kinds of fuels, different phases were recognized, like particle peripheral zones with evidence of Al-U reaction, internal recrystallized zones and bubbles. A very good agreement between code prediction and irradiation results was found. The few discrepancies are due to local, fabrication and irradiation uncertainties, as the presence of U{sub 3}Si phase in U{sub 3}Si{sub 2} particles and effective burnup.« less

40 CFR 52.1167 - EPA-approved Massachusetts State regulations.

Code of Federal Regulations, 2014 CFR

2014-07-01

... coating, wood product surface coating, and flat wood paneling surface coating. 310 CMR 7.04(2) U Fossil... for smoke density instrument removal for certain facilities. 310 CMR 7.04(4)(a) U Fossil Fuel... facilities in that district can apply to burn fossil fuel with an ash content in excess of 9 pct bydry weight...

Characterization of molybdenum particles reinforced Al6082 aluminum matrix composites with improved ductility produced using friction stir processing

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

Selvakumar, S., E-mail: lathaselvam1963@gmail.com

Aluminum matrix composites (AMCs) reinforced with various ceramic particles suffer a loss in ductility. Hard metallic particles can be used as reinforcement to improve ductility. The present investigation focuses on using molybdenum (Mo) as potential reinforcement for Mo(0,6,12 and 18 vol.%)/6082Al AMCs produced using friction stir processing (FSP). Mo particles were successfully retained in the aluminum matrix in its elemental form without any interfacial reaction. A homogenous distribution of Mo particles in the composite was achieved. The distribution was independent upon the region within the stir zone. The grains in the composites were refined considerably due to dynamic recrystallization andmore » pinning effect. The tensile test results showed that Mo particles improved the strength of the composite without compromising on ductility. The fracture surfaces of the composites were characterized with deeply developed dimples confirming appreciable ductility. - Highlights: •Molybdenum particles used as reinforcement for aluminum composites to improve ductility. •Molybdenum particles were retained in elemental form without interfacial reaction. •Homogeneous dispersion of molybdenum particles were observed in the composite. •Molybdenum particles improved tensile strength without major loss in ductility. •Deeply developed dimples on the fracture surfaces confirmed improved ductility.« less

Compatibility evaluation between La 2Mo 2O 9 fast oxide-ion conductor and Ni-based materials

NASA Astrophysics Data System (ADS)

Corbel, Gwenaël; Lacorre, Philippe

2006-05-01

The chemical reactivity of La 2NiO 4+δ and nickel metal or nickel oxide with fast oxide-ion conductor La 2Mo 2O 9 is investigated in the annealing temperature range between 600 and 1000 °C, using room temperature X-ray powder diffraction. Within the La 2NiO 4+δ/La 2Mo 2O 9 system, subsequent reaction is evidenced at relatively low annealing temperature (600 °C), with formation of La 2MoO 6 and NiO. The reaction is complete at 1000 °C. At reverse, no reaction occurs between Ni or NiO and La 2Mo 2O 9 up to 1000 °C. Together with a previous work [G. Corbel, S. Mestiri, P. Lacorre, Solid State Sci. 7 (2005) 1216], the current study shows that Ni-CGO cermets might be chemically and mechanically compatible anode materials to work with LAMOX electrolytes in solid oxide fuel cells.

Interaction Between U-Mo Alloys and Alloys Al-Be

NASA Astrophysics Data System (ADS)

Nikitin, S. N.; Tarasov, B. A.; Shornikov, D. P.

The main objective of the work is the experimental determination of the effect of doping on the kinetics of the interaction of beryllium, aluminum and uranium-molybdenum alloy dispersed in the nuclear fuel. It is shown that an increase in the content of Be in Al leads to a linear decrease in the rate of interaction of the alloy with uranium-molybdenum alloy. Besides AlBe-alloys have higher thermal and mechanical properties than other matrix alloys such as AlSi.

Effect of Molybdenum Incorporation on the Structure and Magnetic Properties of Cobalt Ferrite

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

Orozco, C.; Melendez, A.; Manadhar, S.

Here, we report on the effect of molybdenum (Mo) incorporation on the crystal structure, surface morphology, Mo chemical valence state, and magnetic properties of cobalt ferrite (CoFe 2O 4, referred to CFO). Molybdenum incorporated cobalt ferrite (CoFe 2–xMo xO 4, referred to CFMO) ceramics were prepared by the conventional solid-state reaction method by varying the Mo concentration in the range of x = 0.0–0.3. X-ray diffraction studies indicate that the CFMO materials crystallize in inverse spinel cubic phase. Molybdenum incorporation induced lattice parameter increase from 8.322 to 8.343 Å coupled with a significant increase in density from 5.4 to 5.7more » g/cm 3 was evident in structural analyses. Scanning electron microscopy imaging analyses indicate that the Mo incorporation induces agglomeration of particles leading to larger particle size with increasing x(Mo) values. Detailed X-ray photoelectron spectroscopic (XPS) analyses indicate the increasing Mo content with increasing x from 0.0 to 0.3. XPS confirms that the chemistry of Mo is complex in these CFMO compounds; Mo ions exist in the lower oxidation state (Mo 4+) for higher x while in a mixed chemical valence state (Mo 4+, Mo 5+, Mo 6+) for lower x values. From the temperature-dependent magnetization, the samples show ferrimagnetic behavior including the pristine CFO. From the isothermal magnetization measurements, we find almost 2-fold decrease in coercive field ( H c) from 2143 to 1145 Oe with the increase in Mo doping up to 30%. This doping-dependent H c is consistently observed at all the temperatures measured (4, 100, 200, and 300 K). Furthermore, the saturation magnetization estimated at 4 K and at 1.5 T (from M–H loops) goes through a peak at 92 emu/g (at 15% Mo doping) from 81 emu/g (pristine CFO), and starts decreasing to 79 emu/g (at 30% Mo doping). The results demonstrate that the crystal structure, microstructure, and magnetic properties can be tuned by controlling the Mo-content in the CFMO materials.« less

Effect of Molybdenum Incorporation on the Structure and Magnetic Properties of Cobalt Ferrite

DOE PAGES

Orozco, C.; Melendez, A.; Manadhar, S.; ...

2017-09-26

Here, we report on the effect of molybdenum (Mo) incorporation on the crystal structure, surface morphology, Mo chemical valence state, and magnetic properties of cobalt ferrite (CoFe 2O 4, referred to CFO). Molybdenum incorporated cobalt ferrite (CoFe 2–xMo xO 4, referred to CFMO) ceramics were prepared by the conventional solid-state reaction method by varying the Mo concentration in the range of x = 0.0–0.3. X-ray diffraction studies indicate that the CFMO materials crystallize in inverse spinel cubic phase. Molybdenum incorporation induced lattice parameter increase from 8.322 to 8.343 Å coupled with a significant increase in density from 5.4 to 5.7more » g/cm 3 was evident in structural analyses. Scanning electron microscopy imaging analyses indicate that the Mo incorporation induces agglomeration of particles leading to larger particle size with increasing x(Mo) values. Detailed X-ray photoelectron spectroscopic (XPS) analyses indicate the increasing Mo content with increasing x from 0.0 to 0.3. XPS confirms that the chemistry of Mo is complex in these CFMO compounds; Mo ions exist in the lower oxidation state (Mo 4+) for higher x while in a mixed chemical valence state (Mo 4+, Mo 5+, Mo 6+) for lower x values. From the temperature-dependent magnetization, the samples show ferrimagnetic behavior including the pristine CFO. From the isothermal magnetization measurements, we find almost 2-fold decrease in coercive field ( H c) from 2143 to 1145 Oe with the increase in Mo doping up to 30%. This doping-dependent H c is consistently observed at all the temperatures measured (4, 100, 200, and 300 K). Furthermore, the saturation magnetization estimated at 4 K and at 1.5 T (from M–H loops) goes through a peak at 92 emu/g (at 15% Mo doping) from 81 emu/g (pristine CFO), and starts decreasing to 79 emu/g (at 30% Mo doping). The results demonstrate that the crystal structure, microstructure, and magnetic properties can be tuned by controlling the Mo-content in the CFMO materials.« less

Biological activity of particle exhaust emissions from light-duty diesel engines.

PubMed

Carraro, E; Locatelli, A L; Ferrero, C; Fea, E; Gilli, G

1997-01-01

Whole diesel exhaust has been classified recently as a probable carcinogen, and several genotoxicity studies have found particulate exhaust to be clearly mutagenic. Moreover, genotoxicity of diesel particulate is greatly influenced by fuel nature and type of combustion. In order to obtain an effective environmental pollution control, combustion processes using alternative fuels are being analyzed presently. The goal of this study is to determine whether the installation of exhaust after treatment-devices on two light-duty, exhaust gas recirculation (EGR) valve-equipped diesel engines (1930 cc and 2500 cc) can reduce the mutagenicity associated with particles collected during U.S.A. and European driving cycles. Another interesting object was to compare the ability of alternative biodiesel and conventional diesel fuels to reduce the mutagenic activity associated with collected particles from two light duty diesel engines (both 1930 cc) during the European driving cycle. SOF mutagenicity was assayed using the Salmonella/microsome test (TA 98 and TA 100 strains, +/- S9 fraction). In the first part of our study, the highest mutagenicity was revealed by TA98 strain without enzymatic activation, suggesting a direct-acting mutagenicity prevalence in diesel particulate. The 2500 cc engine revealed twofold mutagenic activity compared with the 1930 cc engine (both EGR valve equipped), whereas an opposite result was found in particulate matter amount. The use of a noncatalytic ceramic trap produced a decrease of particle mutagenic activity in the 2500 cc car, whereas an enhancement in the 1930 cc engine was found. The catalytic converter and the electrostatic filter installed on the 2500 cc engine yielded a light particle amount and an SOF mutagenicity decrease. A greater engine stress was obtained using European driving cycles, which caused the strongest mutagenicity/km compared with the U.S.A. cycles. In the second part of the investigation, even though a small number of assays were available, exhaust emission generation by biodiesel fuel seemed to yield a smaller environmental impact than that of the referenced diesel fuel. The results point out the usefulness of mutagenicity testing in the research of both newer, more efficient automotive aftertreatment devices and less polluting fuels.

UN TRISO Compaction in SiC for FCM Fuel Irradiations

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

Terrani, Kurt A.; Trammell, Michael P.; Kiggans, James O.

2016-11-01

The U.S. Department of Energy Office of Nuclear Energy (DOE-NE) Advanced Fuels Campaign (AFC) is conducting research and development to elevate the technology readiness level of Fully Ceramic Microencapsulated (FCM) fuels, a candidate nuclear fuel with potentially enhanced accident tolerance due to very high fission product retention. One of the early activities in FY17 was to demonstrate production of FCM pellets with uranium nitride TRISO particles. This was carried out in preparation of the larger pellet production campaign in support of the upcoming irradiation testing of this fuel form at INL’s Advanced Test Reactor.

Impact of anaerobic oxidation of methane on the geochemical cycle of redox-sensitive elements at cold-seep sites of the northern South China Sea

NASA Astrophysics Data System (ADS)

Hu, Yu; Feng, Dong; Liang, Qianyong; Xia, Zhen; Chen, Linying; Chen, Duofu

2015-12-01

Cold hydrocarbon seepage is a frequently observed phenomenon along continental margins worldwide. However, little is known about the impact of seeping fluids on the geochemical cycle of redox-sensitive elements. Pore waters from four gravity cores (D-8, D-5, D-7, and D-F) collected from cold-seep sites of the northern South China Sea were analyzed for SO42-, Mg2+, Ca2+, Sr2+, dissolved inorganic carbon (DIC), δ13CDIC, dissolved Fe, Mn, and trace elements (e.g. Mo, U). The sulfate concentration-depth profiles, δ13CDIC values and (ΔDIC+ΔCa2++ΔMg2+)/ΔSO42- ratios suggest that organoclastic sulfate reduction (OSR) is the dominant process in D-8 core. Besides OSR, anaerobic oxidation of methane (AOM) is partially responsible for depletion of sulfate at D-5 and D-7 cores. The sulfate consumption at D-F core is predominantly caused by AOM. The depth of sulfate-methane interface (SMI) and methane diffusive flux of D-F core are calculated to be ~7 m and 0.035 mol m-2 yr-1, respectively. The relatively shallow SMI and high methane flux at D-F core suggest the activity of gas seepage in this region. The concentrations of dissolved uranium (U) were inferred to decrease significantly within the iron reduction zone. It seems that AOM has limited influence on the U geochemical cycling. In contrast, a good correlation between the consumption of sulfate and the removal of molybdenum (Mo) suggests that AOM has a significantly influence on the geochemical cycle of Mo at cold seeps. Accordingly, cold seep environments may serve as an important potential sink in the marine geochemical cycle of Mo.

New constraints on macroscopic compact objects as dark matter candidates from gravitational lensing of type Ia supernovae.

PubMed

Metcalf, R Benton; Silk, Joseph

2007-02-16

We use the distribution, and particularly the skewness, of high redshift type Ia supernovae brightnesses relative to the low redshift sample to constrain the density of macroscopic compact objects (MCOs) in the Universe. The supernova data favor dark matter made of microscopic particles (such as the lightest supersymmetric partner) over MCOs with masses between 10(-2)Mo and 10(10)Mo at 89% confidence. Future data will greatly improve this limit. Combined with other constraints, MCOs larger than one-tenth the mass of Earth (approximately 10(-7)Mo) can be eliminated as the sole constituent of dark matter.

Feasibility studies towards future self-sufficient supply of the 99Mo-99mTc isotopes with Japanese accelerators

PubMed Central

NAKAI, Kozi; TAKAHASHI, Naruto; HATAZAWA, Jun; SHINOHARA, Atsushi; HAYASHI, Yoshihiko; IKEDA, Hayato; KANAI, Yasukazu; WATABE, Tadashi; FUKUDA, Mitsuhiro; HATANAKA, Kichiji

2014-01-01

In order to establish a self-sufficient supply of 99mTc, we studied feasibilities to produce its parent nucleus, 99Mo, using Japanese accelerators. The daughter nucleus, 99mTc, is indispensable for medical diagnosis. 99Mo has so far been imported from abroad, which is separated from fission products generated in nuclear reactors using enriched 235U fuel. We investigated 99mTc production possibilities based on the following three scenarios: (1) 99Mo production by the (n, 2n) reaction by spallation neutrons at the J-PARC injector, LINAC; (2) 99Mo production by the (p, pn) reaction at Ep = 50–80 MeV proton at the RCNP cyclotron; (3) 99mTc direct production with a 20 MeV proton beam from the PET cyclotron. Among these three scenarios, scenario (1) is for a scheme on a global scale, scenario (2) works in a local area, and both cases take a long time for negotiations. Scenario (3) is attractive because we can use nearly 50 PET cyclotrons in Japan for 99mTc production. We here consider both the advantages and disadvantages among the three scenarios by taking account of the Japanese accelerator situation. PMID:25504230

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

Martin, William R.; Lee, John C.; baxter, Alan

Information and measured data from the intial Fort St. Vrain (FSV) high temperature gas reactor core is used to develop a benchmark configuration to validate computational methods for analysis of a full-core, commercial HTR configuration. Large uncertainties in the geometry and composition data for the FSV fuel and core are identified, including: (1) the relative numbers of fuel particles for the four particle types, (2) the distribution of fuel kernel diameters for the four particle types, (3) the Th:U ratio in the initial FSV core, (4) and the buffer thickness for the fissile and fertile particles. Sensitivity studies were performedmore » to assess each of these uncertainties. A number of methods were developed to assist in these studies, including: (1) the automation of MCNP5 input files for FSV using Python scripts, (2) a simple method to verify isotopic loadings in MCNP5 input files, (3) an automated procedure to conduct a coupled MCNP5-RELAP5 analysis for a full-core FSV configuration with thermal-hydraulic feedback, and (4) a methodology for sampling kernel diameters from arbitrary power law and Gaussian PDFs that preserved fuel loading and packing factor constraints. A reference FSV fuel configuration was developed based on having a single diameter kernel for each of the four particle types, preserving known uranium and thorium loadings and packing factor (58%). Three fuel models were developed, based on representing the fuel as a mixture of kernels with two diameters, four diameters, or a continuous range of diameters. The fuel particles were put into a fuel compact using either a lattice-bsed approach or a stochastic packing methodology from RPI, and simulated with MCNP5. The results of the sensitivity studies indicated that the uncertainties in the relative numbers and sizes of fissile and fertile kernels were not important nor were the distributions of kernel diameters within their diameter ranges. The uncertainty in the Th:U ratio in the intial FSV core was found to be important with a crude study. The uncertainty in the TRISO buffer thickness was estimated to be unimportant but the study was not conclusive. FSV fuel compacts and a regular FSV fuel element were analyzed with MCNP5 and compared with predictions using a modified version of HELIOS that is capable of analyzing TRISO fuel configurations. The HELIOS analyses were performed by SSP. The eigenvalue discrepancies between HELIOS and MCNP5 are currently on the order of 1% but these are still being evaluated. Full-core FSV configurations were developed for two initial critical configurations - a cold, clean critical loading and a critical configuration at 70% power. MCNP5 predictions are compared to experimental data and the results are mixed. Analyses were also done for the pulsed neutron experiments that were conducted by GA for the initial FSV core. MCNP5 was used to model these experiments and reasonable agreement with measured results has been observed.« less

3D Field Modifications of Core Neutral Fueling In the EMC3-EIRENE Code

NASA Astrophysics Data System (ADS)

Waters, Ian; Frerichs, Heinke; Schmitz, Oliver; Ahn, Joon-Wook; Canal, Gustavo; Evans, Todd; Feng, Yuehe; Kaye, Stanley; Maingi, Rajesh; Soukhanovskii, Vsevolod

2017-10-01

The application of 3-D magnetic field perturbations to the edge plasmas of tokamaks has long been seen as a viable way to control damaging Edge Localized Modes (ELMs). These 3-D fields have also been correlated with a density drop in the core plasmas of tokamaks; known as `pump-out'. While pump-out is typically explained as the result of enhanced outward transport, degraded fueling of the core may also play a role. By altering the temperature and density of the plasma edge, 3-D fields will impact the distribution function of high energy neutral particles produced through ion-neutral energy exchange processes. Starved of the deeply penetrating neutral source, the core density will decrease. Numerical studies carried out with the EMC3-EIRENE code on National Spherical Tokamak eXperiment-Upgrade (NSTX-U) equilibria show that this change to core fueling by high energy neutrals may be a significant contributor to the overall particle balance in the NSTX-U tokamak: deep core (Ψ < 0.5) fueling from neutral ionization sources is decreased by 40-60% with RMPs. This work was funded by the US Department of Energy under Grant DE-SC0012315.

Effect of High Si Content on U3Si2 Fuel Microstructure

NASA Astrophysics Data System (ADS)

Rosales, Jhonathan; van Rooyen, Isabella J.; Meher, Subhashish; Hoggan, Rita; Parga, Clemente; Harp, Jason

2018-02-01

The development of U3Si2 as an accident-tolerant nuclear fuel has gained research interest because of its promising high uranium density and improved thermal properties. In the present study, three samples of U3Si2 fuel with varying silicon content have been fabricated by a conventional powder metallurgical route. Microstructural characterization via scanning and transmission electron microscopy reveals the presence of other stoichiometry of uranium silicide such as USi and UO2 in both samples. The detailed phase analysis by x-ray diffraction shows the presence of secondary phases, such as USi, U3Si, and UO2. The samples with higher concentrations of silicon content of 7.5 wt.% display additional elemental Si. These samples also possess an increased amount of the USi phase as compared to that in the conventional sample with 7.3 wt.% silicon. The optimization of U3Si2 fuel performance through the understanding of the role of Si content on its microstructure has been discussed.

Method for generating a crystalline {sup 99}MoO{sub 3} product and the isolation {sup 99m}Tc compositions therefrom

DOEpatents

Bennett, R.G.; Christian, J.D.; Kirkham, R.J.; Tranter, T.J.

1998-09-01

An improved method is described for producing {sup 99m}Tc compositions. {sup 100}Mo metal is irradiated with photons in a particle (electron) accelerator to produce {sup 99}Mo metal which is dissolved in a solvent. A solvated {sup 99}Mo product is then dried to generate a supply of {sup 99}MoO{sub 3} crystals. The crystals are thereafter heated at a temperature which will sublimate the crystals and form a gaseous mixture containing vaporized {sup 99m}TcO{sub 3} and vaporized {sup 99m}TcO{sub 2} but will not cause the production of vaporized {sup 99}MoO{sub 3}. The mixture is then combined with an oxidizing gas to generate a gaseous stream containing vaporized {sup 99m}Tc{sub 2}O{sub 7}. Next, the gaseous stream is cooled to a temperature sufficient to convert the vaporized {sup 99m}Tc{sub 2}O{sub 7} into a condensed {sup 99m}Tc-containing product. The product has high purity levels resulting from the use of reduced temperature conditions and ultrafine crystalline {sup 99}MoO{sub 3} starting materials with segregated {sup 99m}Tc compositions therein which avoid the production of vaporized {sup 99}MoO{sub 3} contaminants. 1 fig.

The photoelectronic behaviors of MoO3-loaded ZrO2/carbon cluster nanocomposite materials

NASA Astrophysics Data System (ADS)

Matsui, H.; Ishiko, A.; Karuppuchamy, S.; Hassan, M. A.; Yoshihara, M.

2012-03-01

A novel nano-sized ZrO2/carbon cluster composite materials (Ic's) were successfully obtained by the calcination of ZrCl4/starch complexes I's under an argon atmosphere. Pt- and/or MoO3-loaded ZrO2/carbon clusters composite materials were also prepared by doping Pt and/or MoO3 particles on the surface of Ic's. The surface characterization of the composite materials was carried out using transmission electron microscopy (TEM). The TEM observation of the materials showed the presence of particles with the diameters of a few nanometers, possibly Pt particles, and of 50-100 nm, possibly MoO3 particles, in the matrix. Pt- and/or MoO3-loaded ZrO2/carbon cluster composite materials show the efficient photocatalytic activity under visible light irradiation.

Gasoline-fueled solid oxide fuel cell using MoO2-Based Anode

NASA Astrophysics Data System (ADS)

Hou, Xiaoxue; Marin-Flores, Oscar; Kwon, Byeong Wan; Kim, Jinsoo; Norton, M. Grant; Ha, Su

2014-12-01

This short communication describes the performance of a solid oxide fuel cell (SOFC) fueled by directly feeding premium gasoline to the anode without using external reforming. The novel component of the fuel cell that enables such operation is the mixed conductivity of MoO2-based anode. Using this anode, a fuel cell demonstrating a maximum power density of 31 mW/cm2 at 0.45 V was successfully fabricated. Over a 24 h period of operation, the open cell voltage remained stable at ∼0.92 V. Scanning electron microscopy (SEM) examination of the anode surface pre- and post-testing showed no evidence of coking.

AGR-1 Post Irradiation Examination Final Report

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

Demkowicz, Paul Andrew

The post-irradiation examination (PIE) of the Advanced Gas Reactor (AGR)-1 experiment was a multi-year, collaborative effort between Idaho National Laboratory (INL) and Oak Ridge National Laboratory (ORNL) to study the performance of UCO (uranium carbide, uranium oxide) tristructural isotropic (TRISO) coated particle fuel fabricated in the U.S. and irradiated at the Advanced Test Reactor at INL to a peak burnup of 19.6% fissions per initial metal atom. This work involved a broad array of experiments and analyses to evaluate the level of fission product retention by the fuel particles and compacts (both during irradiation and during post-irradiation heating tests tomore » simulate reactor accident conditions), investigate the kernel and coating layer morphology evolution and the causes of coating failure, and explore the migration of fission products through the coating layers. The results have generally confirmed the excellent performance of the AGR-1 fuel, first indicated during the irradiation by the observation of zero TRISO coated particle failures out of 298,000 particles in the experiment. Overall release of fission products was determined by PIE to have been relatively low during the irradiation. A significant finding was the extremely low levels of cesium released through intact coatings. This was true both during the irradiation and during post-irradiation heating tests to temperatures as high as 1800°C. Post-irradiation safety test fuel performance was generally excellent. Silver release from the particles and compacts during irradiation was often very high. Extensive microanalysis of fuel particles was performed after irradiation and after high-temperature safety testing. The results of particle microanalysis indicate that the UCO fuel is effective at controlling the oxygen partial pressure within the particle and limiting kernel migration. Post-irradiation examination has provided the final body of data that speaks to the quality of the AGR-1 fuel, building on the as-fabricated fuel characterization and irradiation data. In addition to the extensive volume of results generated, the work also resulted in a number of novel analysis techniques and lessons learned that are being applied to the examination of fuel from subsequent TRISO fuel irradiations. This report provides a summary of the results obtained as part of the AGR-1 PIE campaign over its approximately 5-year duration.« less

Design Study of a Modular Gas-Cooled, Closed-Brayton Cycle Reactor for Marine Use

DTIC Science & Technology

1989-06-01

materials in the core and surroundings. To investigate this design point in the marine variant I developed the program HEAT.BAS to perform a one-dimensional...helium as the working fluid. The core is a graphite moderated, epithermal spectrum reactor, using TRISO fuel particles in extruded graphite fuel elements...The fuel is highly enriched U2315 . The containment is shaped in an inverted ’T’ with two sections. The upper section contains the reactor core

Distribution of Pd, Ag & U in the SiC Layer of an Irradiated TRISO Fuel Particle

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

Thomas M. Lillo; Isabella J. van Rooyen

2014-08-01

The distribution of silver, uranium and palladium in the silicon carbide (SiC) layer of an irradiated TRISO fuel particle was studied using samples extracted from the SiC layer using focused ion beam (FIB) techniques. Transmission electron microscopy in conjunction with energy dispersive x-ray spectroscopy was used to identify the presence of the specific elements of interest at grain boundaries, triple junctions and precipitates in the interior of SiC grains. Details on sample fabrication, errors associated with measurements of elemental migration distances and the distances migrated by silver, palladium and uranium in the SiC layer of an irradiated TRISO particle frommore » the AGR-1 program are reported.« less

Analysis of simulated high burnup nuclear fuel by laser induced breakdown spectroscopy

NASA Astrophysics Data System (ADS)

Singh, Manjeet; Sarkar, Arnab; Banerjee, Joydipta; Bhagat, R. K.

2017-06-01

Advanced Heavy Water Reactor (AHWR) grade (Th-U)O2 fuel sample and Simulated High Burn-Up Nuclear Fuels (SIMFUEL) samples mimicking the 28 and 43 GWd/Te irradiated burn-up fuel were studied using laser-induced breakdown spectroscopy (LIBS) setup in a simulated hot-cell environment from a distance of > 1.5 m. Resolution of < 38 pm has been used to record the complex spectra of the SIMFUEL samples. By using spectrum comparison and database matching > 60 emission lines of fission products was identified. Among them only a few emission lines were found to generate calibration curves. The study demonstrates the possibility to investigate impurities at concentrations around hundreds of ppm, rapidly at atmospheric pressure without any sample preparation. The results of Ba and Mo showed the advantage of LIBS analysis over traditional methods involving sample dissolution, which introduces possible elemental loss. Limits of detections (LOD) under Ar atmosphere shows significant improvement, which is shown to be due to the formation of stable plasma.

Calibration study on subchronic inhalation toxicity of man-made vitreous fibers in rats.

PubMed

Bellmann, Bernd; Muhle, Hartwig; Creutzenberg, Otto; Ernst, Heinrich; Müller, Meike; Bernstein, David M; Riego Sintes, Juan M

2003-10-01

This 3-mo inhalation study investigated the biological effects of a special-purpose glass microfiber (E-glass microfiber), the stone wool fiber MMVF21, and a new high-temperature application fiber (calcium-magnesium-silicate fiber, CMS) in Wistar rats. Rats were exposed 6 h/day, 5 days/wk for 3 mo to fiber aerosol concentrations of approximately 15, 50, and 150 fibers/ml (fiber length >20 microm) for E-glass microfiber and MMVF21. For the CMS fiber only the highest exposure concentration was used. During a 3-mo postexposure period, recovery effects were studied. In the highest exposure concentration groups, gravimetric concentrations were 17.2 mg/m3 for E-glass microfiber, 37 mg/m3 for MMVF21, and 49.5 mg/m3 for the CMS fiber. After 3 mo of exposure, lung retention of fibers longer than 20 micro m per lung was 17 x 10(6) for E-glass microfiber, 5.7 x 10(6) for MMVF21, and 0.88 x 10(6) for CMS. After 3 mo of recovery the concentration of the long fiber fraction was decreased to 38.4%, 63.9%, and 3.0% compared to original lung burden for the E-glass microfiber, MMVF21, and CMS, respectively. Biological effects measured included inflammatory and proliferative potential, histopathology lesions, and the persistence of these effects over a recovery period of 3 mo. Generally, observed effects were higher for E-glass microfiber when compared to MMVF21. The following clear dose-dependent effects on E-glass microfiber and MMVF21 exposure were observed as main findings of the study: increase in lung weight, in measured biochemical parameters and polymorphonuclear leukocytes (PMN) in the bronchoalveolar lavage fluid (BALF), in cell proliferation (BrdU-response) of terminal bronchiolar epithelium, and in interstitial fibrosis. The values observed in the proliferation assay on the carcinogenic E-glass microfiber indicate that this assay has an important predictive value with regards to potential carcinogenicity. Surprisingly, for the biosoluble CMS fiber, fibrogenic potential was detected in this study. The results of the CMS exposure group indicate that effects may be dominated by the presence of nonfibrous particles and that fibrosis may not be a predictor of carcinogenic activity of fiber samples, if the fiber preparation contains a significant fraction of nonfibrous particles. In summary, this study demonstrates the importance of fiber dust contamination by granular components. For future subchronic studies a longer posttreatment observation period would be advisable.

NPF MECHANICAL CELL NaK DISPOSAL AND FUME ABATEMENT

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

Rey, G.

Some of the fuels originally scheduled for processing in the nonproduction fuel (NPF) processing program incorporated sodium or sodium- potassium alloy (NaK) as the bonding material between stainless-steel cladding and the uranium or uranium-molybdenum alloy core. Because of the special hazards involved in handling NaK, studies were made to determine safe methods for processing NaK-containing fuels. An underwater NaK dispensing system was installed, and tests were made to determine the characteristics of the NaK-water reaction. The equipment consisted of a dispenser, reaction pan, and off-gas scrubber. After initinl studies, a prototype test was made wherein U-Mo canned slugs containing NaKmore » reservoirs were hack sawed underwater. The studies demonstrated that the NaK reservoirs can be safely deactivated by hack sawing under a submerged hood in a shallow water bath. (W.L.H.)« less

Sequestration of radioactive iodine in silver-palladium phases in commercial spent nuclear fuel

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

Buck, Edgar C.; Mausolf, Edward J.; McNamara, Bruce K.

Radioactive iodine is the Achilles’ heel in the design for the safe geological disposal of spent UO2 nuclear fuel. Iodine’s high solubility and anticipated instant release during waste package compromise jeopardize performance assessment calculations. However, dissolution studies have indicated that the instant release fraction (IRF) of radioiodine (I) does not correlate with increasing fuel burn-up. In fact, there is a peak in the release iodine at around 50-60 Mwd/kgU and with increasing burn-up the instant release of iodine decreases. Detailed electron microscopy analysis of high burn-up fuel (~80 MWd/kgU) has revealed the presence of (Pd,Ag)(I,Br) nano-particles. As UO2 fuels aremore » irradiated, the Ag and Pd content increases, from 239Pu fission, enabling radioiodine to be retained. The occurrence of these phases in nuclear fuels may have significant implications for the long-term behavior of iodine.« less

Temperature dependent elastic properties of γ-phase U – 8 wt% Mo

DOE PAGES

Steiner, M. A.; Garlea, E.; Creasy, J.; ...

2017-12-28

Polycrystalline elastic moduli and stiffness tensor components of γ-phase U – 8 wt% Mo have been determined by resonant ultrasound spectroscopy in the temperature range of 25-650°C. The ambient temperature elastic properties are compared to results measured via other experimental methods and show reasonable agreement, though there is considerable variation of these properties within the literature at both the U – 8 wt% Mo composition and as a function of Mo concentration. The Young’s modulus of U – 8 wt% Mo measured in this study decreases steadily with temperature at a rate that is slower than trends previously observed atmore » similar Mo concentrations, though the difference is not statistically significant. This first measurement of the temperature dependent elastic stiffness tensor of a polycrystalline U-Mo alloy clarifies that the behavior of the Young’s modulus is due to a strongly weakening C 11 polycrystalline stiffness tensor component, along with milder decreases in C 12 and C 44. The unique partially auxetic properties recently predicted for singlecrystalline U-Mo are discussed in regard to their possible impact on the polycrystalline behavior of the alloy.« less

Temperature dependent elastic properties of γ-phase U – 8 wt% Mo

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

Steiner, M. A.; Garlea, E.; Creasy, J.

Polycrystalline elastic moduli and stiffness tensor components of γ-phase U – 8 wt% Mo have been determined by resonant ultrasound spectroscopy in the temperature range of 25-650°C. The ambient temperature elastic properties are compared to results measured via other experimental methods and show reasonable agreement, though there is considerable variation of these properties within the literature at both the U – 8 wt% Mo composition and as a function of Mo concentration. The Young’s modulus of U – 8 wt% Mo measured in this study decreases steadily with temperature at a rate that is slower than trends previously observed atmore » similar Mo concentrations, though the difference is not statistically significant. This first measurement of the temperature dependent elastic stiffness tensor of a polycrystalline U-Mo alloy clarifies that the behavior of the Young’s modulus is due to a strongly weakening C 11 polycrystalline stiffness tensor component, along with milder decreases in C 12 and C 44. The unique partially auxetic properties recently predicted for singlecrystalline U-Mo are discussed in regard to their possible impact on the polycrystalline behavior of the alloy.« less

Urban and rural ultrafine (PM 0.1) particles in the Helsinki area

NASA Astrophysics Data System (ADS)

Pakkanen, Tuomo A.; Kerminen, Veli-Matti; Korhonen, Christina H.; Hillamo, Risto E.; Aarnio, Päivi; Koskentalo, Tarja; Maenhaut, Willy

In June 1996-June 1997 Berner low-pressure impactors were used at an urban and at a rural site in the Helsinki area for sampling ultrafine particles (UFP, PM 0.1). Ten sample pairs, each pair measured simultaneously, were collected in the size range of 0.03-15 μm of particle aerodynamic diameter. More than 40 chemical components were measured. Surprisingly, the average UFP mass concentration was higher at the rural site (520 ng/m 3) than at the urban site (490 ng/m 3). The average chemical composition of UFP was similar at the two sites. The most abundant of the measured components were sulphate (32 and 40 ng/m 3 for the urban and rural sites, respectively), ammonium (22 and 25 ng/m 3), nitrate (4 and 11 ng/m 3) and the Ca 2+ ion (5 and 7 ng/m 3). The most important metals at both sites were Ca, Na, Fe, K and Zn with concentrations between 0.7 and 5 ng/m 3. Of the heavy metals, Ni, V, Cu, and Pb were important with average ultrafine concentrations between about 0.1 and 0.2 ng/m 3. Also the organic anions oxalate (urban 2.1 ng/m 3 and rural 1.9 ng/m 3) and methanesulphonate (1.3 and 1.7 ng/m 3) contributed similarly at both sites. The measured species accounted for only about 15-20% of the total ultrafine mass. The fraction that was not measured includes mainly carbonaceous material and water. It was estimated that the amount of water was about 10% (50 ng/m 3) and that of carbonaceous material about 70% (350 ng/m 3) at both sites. Aitken modes were observed for most components with the average mass mean mode diameters being between about 0.06 and 0.12 μm. The average concentrations in the Aitken mode differed clearly from those in the UFP for several components. The average contribution of ultrafine mass to the fine particle mass (PM 2.5) was about 7% at the urban site and 8.5% at the rural site. At both sites the contribution of ultrafine to fine was especially high for Se, Ag, B, and Ni (10-20%) and at the rural site also for Co (20%), Ca 2+ (16%) and Mo (11%). Enrichment in the ultrafine particles suggests that local sources may exist for these elements. Aitken modes turned out to be useful indicators of local sources for several components. The Aitken modes of Ba, Ca, Mg and Sr were similar in several samples, suggesting a common local combustion source for these elements, possibly traffic exhaust. Co, Fe, Mo and Ni formed another group of elements often having similar Aitken modes, the likely source being combustion of heavy fuel oil.

Use of models to study forest fire behavior

Treesearch

Wallace L. Fons

1961-01-01

The U.S. Forest Service has started a laboratory study with the ultimate objective of determining model laws for fire behavior. The study includes an examination of the effect of such variables as species of wood, density of wood, moisture content, size of fuel particle, spacing, dimensions of fuel bed, wind, and slope on the rate of spread of fire and the partition of...

Cross Section Measurement for the 95Mo(n, {alpha})92Zr Reaction at 4.0, 5.0 and 6.0 MeV

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

Zhang, Guohui; Wu, Hao; Zhang, Jiaguo

2011-01-01

For the {sup 95}Mo(n, {alpha}){sup 92}Zr reaction cross section, there is only one experimental datum in the MeV neutron energy region with large uncertainty. As a result, very large deviations exist in different evaluated nuclear data libraries. This paper report the measurement of cross sections of the {sup 95}Mo(n, {alpha}){sup 92}Zr reaction at En = 4.0, 5.0 and 6.0 MeV. Experiments were performed at the 4.5 MV Van de Graaff of Peking University, China. A twin gridded ionization chamber was used as alpha particle detector and two large area {sup 95}Mo samples placed back to back were adopted. Fast neutronsmore » were produced through the D(d, n){sup 3}He reaction by using a deuterium gas target. A small {sup 238}U fission chamber was adopted for absolute neutron flux determination and a BF{sub 3} long counter was used for neutron flux monitor. Present experimental data are compared with existing evaluations and measurement.« less

On the Mo-Papas equation

NASA Astrophysics Data System (ADS)

Aguirregabiria, J. M.; Chamorro, A.; Valle, M. A.

1982-05-01

A new heuristic derivation of the Mo-Papas equation for charged particles is given. It is shown that this equation cannot be derived for a point particle by closely following Dirac's classical treatment of the problem. The Mo-Papas theory and the Bonnor-Rowe-Marx variable mass dynamics are not compatible.

Emissions of Volatile Particulate Components from Turboshaft Engines running JP-8 and Fischer-Tropsch Fuels

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

Cheng, Mengdawn; Corporan, E.; DeWitt, M.

2009-01-01

Rotating-wing aircraft or helicopters are heavily used by the US military and also a wide range of commercial applications around the world, but emissions data for this class of engines are limited. In this study, we focus on emissions from T700-GE-700 and T700-GE-701C engines; T700 engine was run with military JP-8 and T701C run with both JP-8 and Fischer-Tropsch (FT) fuels. Each engine was run at three engine power settings from the idle to maximum power in sequence. Exhaust particles measured at the engine exhaust plane (EEP) have a peak mobility diameter less than 50nm in all engine power settings.more » At a 4-m downstream location, sulfate/sulfur measurements indicate all particulate sulfur exists practically as sulfate, and the particulate sulfur and sulfate contents increased as the engine power increased. The conversion of sulfur to sulfate was found not to be dependent on engine power setting. Analysis also showed that conversion of sulfur to sulfate was not by the adsorption of sulfur dioxide gas on the soot particles and then subsequently oxidized to form sulfate, but by gas-phase conversion of SO2 via OH or O then subsequently forming H2SO4 and condensing on soot particles. Without the sulfur and aromatic components, use of the FT fuel led to significant reduction of soot emissions as compared to that of the JP-8 fuel producing less number of particles than that of the JP-8 fuel; however, the FT fuel produced much higher number concentrations of particles smaller than 7nm than that of JP-8 in all engine power settings. This indicates non-aromatics components in the FT fuel could have contributed to the enhancement of emissions of particles smaller than 7nm. These small particles are volatile, not observed at the EEP, and may be important in playing a role for the formation of secondary particles in the atmosphere or serving as a site for effective cloud nuclei condensation to occur.« less

Satellite measurements of large-scale air pollution - Measurements of forest fire smoke

NASA Technical Reports Server (NTRS)

Ferrare, Richard A.; Kaufman, Yoram J.; Fraser, Robert S.

1990-01-01

The transport, optical properties, total mass, and removal of smoke produced by forest fires in western Canada during late July and early August 1982 are studied using NOAA 7 AVHRR data. Color composite imagery is produced to track the movement of the smoke over Canada and the U.S. as the smoke traveled thousands of km from the source region. Smoke optical thickness, particle size, and single scattering albedo are computed using radiances measured by AVHRR bands 1 and 2. Results show that smoke optical thickness ranged from less that 0.1 to greater than 3.7 and the geometric mean mass radii ranged from 300 to 900 nm. The smoke single scattering albedo ranged from 0.9 to nearly 1.0. The total smoke mass over the eastern U.S. ranged from 0.1 to 0.5 Tg, which is close to the 0.5 Tg estimated from the forest fuel content. The smoke lifetime is estimated to be between 15 and 20 days.

Theoretical Model for Volume Fraction of UC, 235U Enrichment, and Effective Density of Final U 10Mo Alloy

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

Devaraj, Arun; Prabhakaran, Ramprashad; Joshi, Vineet V.

2016-04-12

The purpose of this document is to provide a theoretical framework for (1) estimating uranium carbide (UC) volume fraction in a final alloy of uranium with 10 weight percent molybdenum (U-10Mo) as a function of final alloy carbon concentration, and (2) estimating effective 235U enrichment in the U-10Mo matrix after accounting for loss of 235U in forming UC. This report will also serve as a theoretical baseline for effective density of as-cast low-enriched U-10Mo alloy. Therefore, this report will serve as the baseline for quality control of final alloy carbon content

National Dam Safety Program. Lipps Lake Dam (MO 30214). Mississippi - Kaskaskia - St. Louis Basin, Cape Girardeau County, Missouri. Phase I Inspection Report.

DTIC Science & Technology

1980-10-01

AD-A105 988 HOSKINSEWESTERN-SONOER EGGER INC LINCOLN NEM F/S 13/13 NATIONAL DA -M SAFETY PROGRAM. LI PS LAKE DAM (MO 3021 ). MISS! SS7 - TC(U...COMPLETING FORM i. REPORT NUMBER 12. GOVT ACCESSION NO. 3. RECIPIENT’S CATALOG NIOMBER 4. TITLE (and Subtitle) 5. TYPE OF REPORT & PERIOD COVERED Phase I Dam

The interaction of molecular hydrogen with α-radiolytic oxidants on a (U,Pu)O2 surface

NASA Astrophysics Data System (ADS)

Bauhn, Lovisa; Hansson, Niklas; Ekberg, Christian; Fors, Patrik; Delville, Rémi; Spahiu, Kastriot

2018-07-01

In order to assess the impact of α-radiolysis of water on the oxidative dissolution of spent fuel, an un-irradiated, annealed MOX fuel pellet with high content of Pu (∼24 wt%), and a specific α-activity of 4.96 GBq/gMOX, was leached in carbonate-containing solutions of low ionic strength. The high Pu content in the pellet stabilizes the (U,Pu)O2(s) matrix towards oxidative dissolution, whereas the α-decays emitted from the surface are expected to produce ∼3.6 × 10-7 mol H2O2/day, contributing to the oxidative dissolution of the pellet. Two sets of leaching tests were conducted under different redox conditions: Ar gas atmosphere and deuterium gas atmosphere. A relatively slow increase of the U and Pu concentrations was observed in the Ar case, with U concentrations increasing from 1·10-6 M after 1 h to ∼7 × 10-5 M after 58 days. Leaching under an atmosphere starting at 1 MPa deuterium gas was undertaken in order to evaluate any effect of dissolved hydrogen on the radiolytic dissolution of the pellet, as well as to investigate any potential recombination of the α-radiolytic products with dissolved deuterium. For the latter purpose, isotopic analysis of the D/H content was carried out on solution samples taken during the leaching. Despite the continuous production of radiolytic oxidants, the concentrations of U and Pu remained quite constant at the level of ∼3 × 10-8 M during the first 30 days, i.e. as long as the deuterium pressure remained higher than 0.8 MPa. These data rule out any oxidative dissolution of the pellet during the first month. The un-irradiated MOX fuel does not contain metallic ε-particles, hence it is mainly the interaction of radiolytic oxidants and dissolved deuterium with the surface of the mixed actinide oxide that causes the neutralization of the oxidants. This conclusion is supported by the steadily increasing levels of HDO measured in the leachate samples.

Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

NASA Astrophysics Data System (ADS)

González-Robles, E.; Serrano-Purroy, D.; Sureda, R.; Casas, I.; de Pablo, J.

2015-10-01

The denominated instant release fraction (IRF) is considered in performance assessment (PA) exercises to govern the dose that could arise from the repository. A conservative definition of IRF comprises the total inventory of radionuclides located in the gap, fractures, and the grain boundaries and, if present, in the high burn-up structure (HBS). The values calculated from this theoretical approach correspond to an upper limit that likely does not correspond to what it will be expected to be instantaneously released in the real system. Trying to ascertain this IRF from an experimental point of view, static leaching experiments have been carried out with two commercial UO2 spent nuclear fuels (SNF): one from a pressurized water reactor (PWR), labelled PWR, with an average burn-up (BU) of 52 MWd/kgU and fission gas release (FGR) of 23.1%, and one from a boiling water reactor (BWR), labelled BWR, with an average BU of and 53 MWd/kgU and FGR of 3.9%. One sample of each SNF, consisting of fuel and cladding, has been leached in bicarbonate water during one year under oxidizing conditions at room temperature (25 ± 5)°C. The behaviour of the concentration measured in solution can be divided in two according to the release rate. All radionuclides presented an initial release rate that after some days levels down to a slower second one, which remains constant until the end of the experiment. Cumulative fraction of inventory in aqueous phase (FIAPc) values has been calculated. Results show faster release in the case of the PWR SNF. In both cases Np, Pu, Am, Cm, Y, Tc, La and Nd dissolve congruently with U, while dissolution of Zr, Ru and Rh is slower. Rb, Sr, Cs and Mo, dissolve faster than U. The IRF of Cs at 10 and 200 days has been calculated, being (3.10 ± 0.62) and (3.66 ± 0.73) for PWR fuel, and (0.35 ± 0.07) and (0.51 ± 0.10) for BWR fuel.

Assessment of Possible Cycle Lengths for Fully-Ceramic Micro-Encapsulated Fuel-Based Light Water Reactor Concepts

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

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

2012-04-01

The tri-isotropic (TRISO) fuel developed for High Temperature reactors is known for its extraordinary fission product retention capabilities [1]. Recently, the possibility of extending the use of TRISO particle fuel to Light Water Reactor (LWR) technology, and perhaps other reactor concepts, has received significant attention [2]. The Deep Burn project [3] currently focuses on once-through burning of transuranic fissile and fissionable isotopes (TRU) in LWRs. The fuel form for this purpose is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the TRISO fuel particle design from high temperature reactor technology, but uses SiC as a matrix material rather thanmore » graphite. In addition, FCM fuel may also use a cladding made of a variety of possible material, again including SiC as an admissible choice. The FCM fuel used in the Deep Burn (DB) project showed promising results in terms of fission product retention at high burnup values and during high-temperature transients. In the case of DB applications, the fuel loading within a TRISO particle is constituted entirely of fissile or fissionable isotopes. Consequently, the fuel was shown to be capable of achieving reasonable burnup levels and cycle lengths, especially in the case of mixed cores (with coexisting DB and regular LWR UO2 fuels). In contrast, as shown below, the use of UO2-only FCM fuel in a LWR results in considerably shorter cycle length when compared to current-generation ordinary LWR designs. Indeed, the constraint of limited space availability for heavy metal loading within the TRISO particles of FCM fuel and the constraint of low (i.e., below 20 w/0) 235U enrichment combine to result in shorter cycle lengths compared to ordinary LWRs if typical LWR power densities are also assumed and if typical TRISO particle dimensions and UO2 kernels are specified. The primary focus of this summary is on using TRISO particles with up to 20 w/0 enriched uranium kernels loaded in Pressurized Water Reactor (PWR) assemblies. In addition to consideration of this 'naive' use of TRISO fuel in LWRs, several refined options are briefly examined and others are identified for further consideration including the use of advanced, high density fuel forms and larger kernel diameters and TRISO packing fractions. The combination of 800 {micro}m diameter kernels of 20% enriched UN and 50% TRISO packing fraction yielded reactivity sufficient to achieve comparable burnup to present-day PWR fuel.« less

High Fragmentation Steel Production Process

DTIC Science & Technology

1984-01-01

J/ FTA c« ;« MO G SO KM s s P WS W-U Hi ; T 14 434 CASK G S3 K 11 ma WM MM MM ACTS 1 TC*4 U S7« ill GC 135 V M NTA «M FT...relative feed range 2nd digit -relative force range FMd 1 Very Low Fore* t 2 Low 2 3 Medium Low 3 4 Medium 4 5 Medium 5 6 Medium High 6 7 Medium

Feasibility of rotating fluidized bed reactor for rocket propulsion

NASA Technical Reports Server (NTRS)

Ludewig, H.; Manning, A. J.; Raseman, C. J.

1974-01-01

The rotating fluidized bed reactor concept is outlined, and its application to rocket propulsion is discussed. Experimental results obtained indicate that minimum fluidization correlations commonly in use for 1-g beds can also be applied to multiple-g beds. It was found that for a low thrust system (20,000 lbf) the fuel particle size and/or particle stress play a limiting role on performance. The superiority of U-233 as a fuel for this type of rocket engine is clearly demonstrated in the analysis. The maximum thrust/weight ratio for a 90,000N thrust engine was found to be approximately 65N/kg.

Neutron Resonance Densitometry for Particle-like Debris of Melted Fuel

NASA Astrophysics Data System (ADS)

Harada, H.; Kitatani, F.; Koizumi, M.; Takamine, J.; Kureta, M.; Tsutiya, H.; Iimura, H.; Seya, M.; Becker, B.; Kopecky, S.; Schillebeeckx, P.

2014-04-01

Neutron Resonance Densitometry (NRD) is proposed for the quantification of nuclear materials in particle-like debris of melted fuel from the reactors of the Fukushima Daiichi nuclear power plant. The method is based on a combination of neutron resonance transmission analysis (NRTA) and neutron resonance capture analysis (NRCA). It uses the neutron time-of-flight (TOF) technique with a pulsed white neutron source and a neutron flight path as short as 5 m. The spectrometer for NRCA is made of LaBr3(Ce) detectors. The achievable uncertainty due to only counting statistics is less than 1 % to determine Pu and U isotopes.

7 CFR 2.73 - Director, Office of Energy Policy and New Uses.

Code of Federal Regulations, 2010 CFR

2010-01-01

... competitive biodiesel fuel education grants program (7 U.S.C. 8106). (9) Implement a memorandum of understanding with the Secretary of Energy regarding cooperation in the application of hydrogen and fuel cell...

7 CFR 2.73 - Director, Office of Energy Policy and New Uses.

Code of Federal Regulations, 2011 CFR

2011-01-01

... competitive biodiesel fuel education grants program (7 U.S.C. 8106). (9) Implement a memorandum of understanding with the Secretary of Energy regarding cooperation in the application of hydrogen and fuel cell...

7 CFR 2.73 - Director, Office of Energy Policy and New Uses.

Code of Federal Regulations, 2014 CFR

2014-01-01

... competitive biodiesel fuel education grants program (7 U.S.C. 8106). (9) Implement a memorandum of understanding with the Secretary of Energy regarding cooperation in the application of hydrogen and fuel cell...

7 CFR 2.73 - Director, Office of Energy Policy and New Uses.

Code of Federal Regulations, 2013 CFR

2013-01-01

... competitive biodiesel fuel education grants program (7 U.S.C. 8106). (9) Implement a memorandum of understanding with the Secretary of Energy regarding cooperation in the application of hydrogen and fuel cell...

7 CFR 2.73 - Director, Office of Energy Policy and New Uses.

Code of Federal Regulations, 2012 CFR

2012-01-01

... competitive biodiesel fuel education grants program (7 U.S.C. 8106). (9) Implement a memorandum of understanding with the Secretary of Energy regarding cooperation in the application of hydrogen and fuel cell...

Estimate of radiation release from MIT reactor with un-finned LEU core during Maximum Hypothetical Accident

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

Sun, Kaichao; Hu, Lin-wen; Newton, Thomas

2017-05-01

The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. At 6 MW, it delivers neutron flux and energy spectrum comparable to light water reactor (LWR) power reactors in a compact core using highly enriched uranium (HEU) fuel. In the framework of nonproliferation policy, the international community aims to minimize the use of HEU in civilian facilities. Within this context, research and test reactors have started a program to convert HEU fuel to low enriched uranium (LEU) fuel. A new type of LEU fuel basedmore » on a high density alloy of uranium and molybdenum (U-10Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MITR. The current study focuses on the impacts of MITR Maximum Hypothetical Accident (MHA), which is also the Design Basis Accident (DBA), with LEU fuel. The MHA for the MITR is postulated to be a coolant flow blockage in the fuel element that contains the hottest fuel plate. It is assumed that the entire active portion of five fuel plates melts. The analysis shows that, within a 2-h period and by considering all the possible radiation sources and dose pathways, the overall off-site dose is 302.1 mrem (1 rem ¼ 0.01 Sv) Total Effective Dose Equivalent (TEDE) at 8 m exclusion area boundary (EAB) and a higher dose of 392.8 mrem TEDE is found at 21 m EAB. In all cases the dose remains below the 500 mrem total TEDE limit goal based on NUREG-1537 guidelines.« less

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

Silva, Chinthaka M.; Hunt, Rodney Dale; Snead, Lance Lewis

Uranium mononitride (UN) is important as a nuclear fuel. Fabrication of UN in its microspherical form also has its own merits since the advent of the concept of accident-tolerant fuel, where UN is being considered as a potential fuel in the form of TRISO particles. But, not many processes have been well established to synthesize kernels of UN. Therefore, a process for synthesis of microspherical UN with a minimum amount of carbon is discussed herein. First, a series of single-phased microspheres of uranium sesquinitride (U 2N 3) were synthesized by nitridation of UO 2+C microspheres at a few different temperatures.more » Resulting microspheres were of low-density U 2N 3 and decomposed into low-density UN. The variation of density of the synthesized sesquinitrides as a function of its chemical composition indicated the presence of extra (interstitial) nitrogen atoms corresponding to its hyperstoichiometry, which is normally indicated as α-U 2N 3. Average grain sizes of both U 2N 3 and UN varied in a range of 1–2.5 μm. In addition, these had a considerably large amount of pore spacing, indicating the potential sinterability of UN toward its use as a nuclear fuel.« less

Formation of Ultrafine Metal Particles by Gas-Evaporation VI. Bcc Metals, Fe, V, Nb, Ta, Cr, Mo and W

NASA Astrophysics Data System (ADS)

Saito, Yahachi; Mihama, Kazuhiro; Uyeda, Ryozi

1980-09-01

The crystal structures and habits of bcc metal particles have been investigated systematically by electron microscopy. The habits for the bcc structure are rhombic dodecahedra truncated by six {100} faces with various degrees of truncation from 0 to 100%. The truncation degree for Fe and V particles grown in the intermediate zone of a metal smoke is in good agreement with that for the Wulff polyhedron expected from the surface energies calculated for {110} and {100} faces. Particles of Cr, Mo and W have the A-15 type structure besides the ordinary bcc structure. The present results support the hypothesis that the A-15 type structure is stable when the particle size is small. The habits for the A-15 type structure are rhombic dodecahedra (Cr), {211} icositetrahedra (Cr and Mo) and rounded cubes (Mo and W).

Synthesis of phase-pure U 2N 3 microspheres and its decomposition into UN

DOE PAGES

Silva, Chinthaka M.; Hunt, Rodney Dale; Snead, Lance Lewis; ...

2014-12-12

Uranium mononitride (UN) is important as a nuclear fuel. Fabrication of UN in its microspherical form also has its own merits since the advent of the concept of accident-tolerant fuel, where UN is being considered as a potential fuel in the form of TRISO particles. But, not many processes have been well established to synthesize kernels of UN. Therefore, a process for synthesis of microspherical UN with a minimum amount of carbon is discussed herein. First, a series of single-phased microspheres of uranium sesquinitride (U 2N 3) were synthesized by nitridation of UO 2+C microspheres at a few different temperatures.more » Resulting microspheres were of low-density U 2N 3 and decomposed into low-density UN. The variation of density of the synthesized sesquinitrides as a function of its chemical composition indicated the presence of extra (interstitial) nitrogen atoms corresponding to its hyperstoichiometry, which is normally indicated as α-U 2N 3. Average grain sizes of both U 2N 3 and UN varied in a range of 1–2.5 μm. In addition, these had a considerably large amount of pore spacing, indicating the potential sinterability of UN toward its use as a nuclear fuel.« less

RERTR-12 Insertion 2 Irradiation Summary Report

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

D. M. Perez; G. S. Chang; D. M. Wachs

2012-09-01

The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-12 was designed to provide comprehensive information on the performance of uranium-molybdenum (U-Mo) based monolithic fuels for research reactor applications.1 RERTR-12 insertion 2 includes the capsules irradiated during the last three irradiation cycles. These capsules include Z, Y1, Y2 and Y3 type capsules. The following report summarizes the life of the RERTR-12 insertion 2 experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.

The Government of Islamic Republic of Afghanistan’s Controls Over the Contract Management Process for U.S. Direct Assistance Need Improvement

DTIC Science & Technology

2015-02-25

develop, award, execute, or monitor individual contracts funded with U.S. direct assistance. This occurred because the Ministry of Finance ( MoF ...trained staff to collaborate with the Ministry of Finance ( MoF ), MoD, and MoI for budgeting, acquisition planning, procurement, financial management...and contract management and oversight. Finally, CSTC-A must ensure MoF , MoD, and MoI establish standard operating procedures and maintain adequate

One-Step Synthesis of Water-Soluble MoS2 Quantum Dots via a Hydrothermal Method as a Fluorescent Probe for Hyaluronidase Detection.

PubMed

Gu, Wei; Yan, Yinghan; Zhang, Cuiling; Ding, Caiping; Xian, Yuezhong

2016-05-11

In this work, a bottom-up strategy is developed to synthesize water-soluble molybdenum disulfide quantum dots (MoS2 QDs) through a simple, one-step hydrothermal method using ammonium tetrathiomolybdate [(NH4)2MoS4] as the precursor and hydrazine hydrate as the reducing agent. The as-synthesized MoS2 QDs are few-layered with a narrow size distribution, and the average diameter is about 2.8 nm. The resultant QDs show excitation-dependent blue fluorescence due to the polydispersity of the QDs. Moreover, the fluorescence can be quenched by hyaluronic acid (HA)-functionalized gold nanoparticles through a photoinduced electron-transfer mechanism. Hyaluronidase (HAase), an endoglucosidase, can cleave HA into proangiogenic fragments and lead to the aggregation of gold nanoparticles. As a result, the electron transfer is blocked and fluorescence is recovered. On the basis of this principle, a novel fluorescence sensor for HAase is developed with a linear range from 1 to 50 U/mL and a detection limit of 0.7 U/mL.

Prime Contractors with Awards Over $25,000 by Name, Location, and Contract Number, Fiscal Year 1992 (Daubert Industries, Inc.-Gaylord Industries, Inc.)

DTIC Science & Technology

1993-01-01

a0 m~ ar (0 - a 0 a4 04 H50 4 a 4 C-1 am a C3 L) 0ý an 0 a n c Za4 00 C.ým 4Lo j I N O Nol 0IS)0 cl r- m - anNe t"I ý ý Ll2 an N -z Do~ acca -fN- 0a’J...c to aQ Coo," I "o o) " ’ mC, ,I e nnI to a) ammc)oc Il I I L) o c wo x I Ca u m2 a. U r-7f cc ~ nl.< C0C) -*4~-0na-ao " 0 ELI ao (v acCa ’’ a~al~ nal...Fw IE~ 1. E F1 F4 o F. c4I F 4 F OF. F7 )I - F ~-4 .F 4 F 41) w 10 A F F W)( o4 v7- m ’mo7 l T =444 45 o o .7 c> S . o Lc4 mo - 04 NNr a)0 lo 7 1 7 o

Naturally Occurring versus Anthropogenic Sources of Elevated Molybdenum in Groundwater: Evidence for Geogenic Contamination from Southeast Wisconsin, United States.

PubMed

Harkness, Jennifer S; Darrah, Thomas H; Moore, Myles T; Whyte, Colin J; Mathewson, Paul D; Cook, Tyson; Vengosh, Avner

2017-11-07

Molybdenum (Mo) is an essential trace nutrient but has negative health effects at high concentrations. Groundwater typically has low Mo (<2 μg/L), and elevated levels are associated with anthropogenic contamination, although geogenic sources have also been reported. Coal combustion residues (CCRs) are enriched in Mo, and thus present a potential anthropogenic contamination source. Here, we use diagnostic geochemical tracers combined with groundwater residence time indicators to investigate the sources of Mo in drinking-water wells from shallow aquifers in a region of widespread CCR disposal in southeastern Wisconsin. Samples from drinking-water wells were collected in areas near and away from known CCR disposal sites, and analyzed for Mo and inorganic geochemistry indicators, including boron and strontium isotope ratios, along with groundwater tritium-helium and radiogenic 4 He in-growth age-dating techniques. Mo concentrations ranged from <1 to 149 μg/L. Concentrations exceeding the U.S. Environmental Protection Agency health advisory of 40 μg/L were found in deeper, older groundwater (mean residence time >300 y). The B (δ 11 B = 22.9 ± 3.5‰) and Sr ( 87 Sr/ 86 Sr = 0.70923 ± 0.00024) isotope ratios were not consistent with the expected isotope fingerprints of CCRs, but rather mimic the compositions of local lithologies. The isotope signatures combined with mean groundwater residence times of more than 300 years for groundwater with high Mo concentrations support a geogenic source of Mo to the groundwater, rather than CCR-induced contamination. This study demonstrates the utility of a multi-isotope approach to distinguish between fossil fuel-related and natural sources of groundwater contamination.

Long-Term Fuel-Specific NO x and Particle Emission Trends for In-Use Heavy-Duty Vehicles in California.

PubMed

Haugen, Molly J; Bishop, Gary A

2018-05-15

Two California heavy-duty fleets have been measured in 2013, 2015, and 2017 using the On-Road Heavy-Duty Measurement System. The Port of Los Angeles drayage fleet has increased in age by 3.3 model years (4.2-7.5 years old) since 2013, with little fleet turnover. Large increases in fuel-specific particle emissions (PM) observed in 2015 were reversed in 2017, returning to near 2013 levels, suggesting repairs and or removal of high emitting vehicles. Fuel-specific oxides of nitrogen (NO x ) emissions of this fleet have increased, and NO x after-treatment systems do not appear to perform ideally in this setting. At the Cottonwood weigh station in northern California, the fleet age has declined (7.8 to 6 years old) since 2013 due to fleet turnover, significantly lowering the average fuel-specific emissions for PM (-87%), black carbon (-76%), and particle number (-64%). Installations of retrofit-diesel particulate filters in model year 2007 and older vehicles have further decreased particle emissions. Cottonwood fleet fuel-specific NO x emissions have decreased slightly (-8%) during this period; however, newer technology vehicles with selective catalytic reduction systems (SCR) promise an additional factor of 4-5 further reductions in the long-haul fleet emissions as California transitions to an all SCR-equipped fleet.

Department of Defense Civilian Personnel Manual

DTIC Science & Technology

1996-12-01

Q Fri . Q 22. Date Mo. Day Yr. notice received U 114 I 95 I 23. Date Mo. Day Yr. stopped «-I *-m- work I 1...DSun. (jp Mon. £] Tues. ©Wed. ElThurs. K2 Fri . DSat. . Name and address of physician first providing medical care (include city, state...Regular ^ hours From: 7 :0OLJ Pm- • a.m. To: 3 :3GQp-m. 22. Regular work schedule (ZlSun. Mon. ]Tues. fJWed. ^Thurs. 13 Fri . Qs

A thermodynamic approach for advanced fuels of gas-cooled reactors

NASA Astrophysics Data System (ADS)

Guéneau, C.; Chatain, S.; Gossé, S.; Rado, C.; Rapaud, O.; Lechelle, J.; Dumas, J. C.; Chatillon, C.

2005-09-01

For both high temperature reactor (HTR) and gas cooled fast reactor (GFR) systems, the high operating temperature in normal and accidental conditions necessitates the assessment of the thermodynamic data and associated phase diagrams for the complex system constituted of the fuel kernel, the inert materials and the fission products. A classical CALPHAD approach, coupling experiments and thermodynamic calculations, is proposed. Some examples of studies are presented leading with the CO and CO 2 gas formation during the chemical interaction of [UO 2± x/C] in the HTR particle, and the chemical compatibility of the couples [UN/SiC], [(U, Pu)N/SiC], [(U, Pu)N/TiN] for the GFR system. A project of constitution of a thermodynamic database for advanced fuels of gas-cooled reactors is proposed.

Tailpipe emissions from gasoline direct injection (GDI) and port fuel injection (PFI) vehicles at both low and high ambient temperatures.

PubMed

Zhu, Rencheng; Hu, Jingnan; Bao, Xiaofeng; He, Liqiang; Lai, Yitu; Zu, Lei; Li, Yufei; Su, Sheng

2016-09-01

Vehicle emissions are greatly influenced by various factors that are related to engine technology and driving conditions. Only the fuel injection method and ambient temperature are investigated in this research. Regulated gaseous and particulate matter (PM) emissions from two advanced gasoline-fueled vehicles, one with direct fuel injection (GDI) and the other with port fuel injection (PFI), are tested with conventional gasoline and ethanol-blended gasoline (E10) at both -7 °C and 30 °C. The total particle number (PN) concentrations and size distributions are monitored with an Electrical Low Pressure Impactor (ELPI(+)). The solid PN concentrations are measured with a condensation particle counter (CPC) after removing volatile matters through the particle measurement program (PMP) system. The results indicate that decreasing the ambient temperature from 30 °C to -7 °C significantly increases the fuel consumption and all measured emissions except for NOx. The GDI vehicle exhibits lower fuel consumption than the PFI vehicle but emits more total hydrocarbons (THC), PM mass and solid PN emissions at 30 °C. The adaptability of GDI technology appears to be better than that of PFI technology at low ambient temperature. For example, the CO, THC and PM mass emission factors of the PFI vehicle are higher than those of the GDI vehicle and the solid PN emission factors are comparable in the cold-start tests at -7 °C. Specifically, during start-up the particulate matter emissions of the PFI are much higher than the GDI. In most cases, the geometric mean diameter (GMD) of the accumulation mode particles is 58-86 nm for both vehicles, and the GMD of the nucleation mode particles is 10-20 nm. The results suggest that the gaseous and particulate emissions from the PFI vehicle should not be neglected compared to those from the GDI vehicle especially in a cold environment. Copyright © 2016 Elsevier Ltd. All rights reserved.

Electron tomography and fractal aspects of MoS2 and MoS2/Co spheres.

PubMed

Ramos, Manuel; Galindo-Hernández, Félix; Arslan, Ilke; Sanders, Toby; Domínguez, José Manuel

2017-09-26

A study was made by a combination of 3D electron tomography reconstruction methods and N 2 adsorption for determining the fractal dimension for nanometric MoS 2 and MoS 2 /Co catalyst particles. DFT methods including Neimarke-Kiselev's method allowed to determine the particle porosity and fractal arrays at the atomic scale for the S-Mo-S(Co) 2D- layers that conform the spherically shaped catalyst particles. A structural and textural correlation was sought by further characterization performed by x-ray Rietveld refinement and Radial Distribution Function (RDF) methods, electron density maps, computational density functional theory methods and nitrogen adsorption methods altogether, for studying the structural and textural features of spherical MoS 2 and MoS 2 /Co particles. Neimark-Kiselev's equations afforded the evaluation of a pore volume variation from 10 to 110 cm 3 /g by cobalt insertion in the MoS 2 crystallographic lattice, which induces the formation of cavities and throats in between of less than 29 nm, with a curvature radius r k  < 14.4 nm; typical large needle-like arrays having 20 2D layers units correspond to a model consisting of smooth surfaces within these cavities. Decreasing D P , D B , D I and D M values occur when Co atoms are present in the MoS 2 laminates, which promote the formation of smoother edges and denser surfaces that have an influence on the catalytic properties of the S-Mo-S(Co) system.

Emissions characterization of residential wood-fired hydronic heater technologies

NASA Astrophysics Data System (ADS)

Kinsey, John S.; Touati, Abderrahmane; Yelverton, Tiffany L. B.; Aurell, Johanna; Cho, Seung-Hyun; Linak, William P.; Gullett, Brian K.

2012-12-01

Residential wood-fired hydronic heaters (RWHHs) can negatively impact the local ambient air quality and thus are an environmental concern in wood burning areas of the U. S. Only a few studies have been conducted which characterize the emissions from RWHHs. To address the lack of emissions data, a study was conducted on four appliances of differing design using multiple fuel types to determine their thermal, boiler, and combustion efficiency as well as the emissions of carbon dioxide (CO2), carbon monoxide (CO), total hydrocarbons (THC), nitrous oxide (N2O), methane (CH4), total particulate matter (PM) mass, and particle number as well as particle size distribution (PSD). Three of these appliances were fired with split-log cordwood with the fourth unit using hardwood pellets. The measured thermal efficiencies for the appliances tested varied from 22 to 44% and the combustion efficiencies from 81 to 98%. Depending on appliance and fuel type, the emission factors ranged from about 1300 to 1800 g kg-1 dry fuel for CO2, 8-190 g kg-1 dry fuel for CO, <1-54 g kg-1 dry fuel for THC and 6-120 mg kg-1 for N2O. For the particle phase pollutants, the PM mass emission factors ranged from 0.31 to 47 g kg-1 dry fuel and the PM number emission factors from 8.5 × 1010 to 2.4 × 1014 particles kg-1 dry fuel, also depending on the appliance and fuel tested. The PSD for all four appliances indicated a well established accumulation mode with evidence of a nucleation mode present for Appliances A and B. The average median aerodynamic particle diameters observed for the four appliances ranged from 84 to 187 nm while burning red oak or pellets. In general, the pellet-burning appliance had the highest overall operating efficiency and lowest emissions of the four units tested.

Bimodal collagen fibril diameter distributions direct age-related variations in tendon resilience and resistance to rupture

PubMed Central

Holmes, D. F.; Lu, Y.; Purslow, P. P.; Kadler, K. E.; Bechet, D.; Wess, T. J.

2012-01-01

Scaling relationships have been formulated to investigate the influence of collagen fibril diameter (D) on age-related variations in the strain energy density of tendon. Transmission electron microscopy was used to quantify D in tail tendon from 1.7- to 35.3-mo-old (C57BL/6) male mice. Frequency histograms of D for all age groups were modeled as two normally distributed subpopulations with smaller (DD1) and larger (DD2) mean Ds, respectively. Both DD1 and DD2 increase from 1.6 to 4.0 mo but decrease thereafter. From tensile tests to rupture, two strain energy densities were calculated: 1) uE [from initial loading until the yield stress (σY)], which contributes primarily to tendon resilience, and 2) uF [from σY through the maximum stress (σU) until rupture], which relates primarily to resistance of the tendons to rupture. As measured by the normalized strain energy densities uE/σY and uF/σU, both the resilience and resistance to rupture increase with increasing age and peak at 23.0 and 4.0 mo, respectively, before decreasing thereafter. Multiple regression analysis reveals that increases in uE/σY (resilience energy) are associated with decreases in DD1 and increases in DD2, whereas uF/σU (rupture energy) is associated with increases in DD1 alone. These findings support a model where age-related variations in tendon resilience and resistance to rupture can be directed by subtle changes in the bimodal distribution of Ds. PMID:22837169

U.S.-Australia Civilian Nuclear Cooperation: Issues for Congress

DTIC Science & Technology

2010-09-30

7 Uranium Mining and Milling ................................................................................................8...cycle begins with mining uranium ore and upgrading it to yellowcake. Because naturally occurring uranium lacks sufficient fissile 235U to make fuel for...enrichment, and finally fabrication into fuel elements. Australia exports its uranium after the mining and milling stage. Commercial enrichment services

Development of multilayer perceptron networks for isothermal time temperature transformation prediction of U-Mo-X alloys

NASA Astrophysics Data System (ADS)

Johns, Jesse M.; Burkes, Douglas

2017-07-01

In this work, a multilayered perceptron (MLP) network is used to develop predictive isothermal time-temperature-transformation (TTT) models covering a range of U-Mo binary and ternary alloys. The selected ternary alloys for model development are U-Mo-Ru, U-Mo-Nb, U-Mo-Zr, U-Mo-Cr, and U-Mo-Re. These model's ability to predict 'novel' U-Mo alloys is shown quite well despite the discrepancies between literature sources for similar alloys which likely arise from different thermal-mechanical processing conditions. These models are developed with the primary purpose of informing experimental decisions. Additional experimental insight is necessary in order to reduce the number of experiments required to isolate ideal alloys. These models allow test planners to evaluate areas of experimental interest; once initial tests are conducted, the model can be updated and further improve follow-on testing decisions. The model also improves analysis capabilities by reducing the number of data points necessary from any particular test. For example, if one or two isotherms are measured during a test, the model can construct the rest of the TTT curve over a wide range of temperature and time. This modeling capability reduces the cost of experiments while also improving the value of the results from the tests. The reduced costs could result in improved material characterization and therefore improved fundamental understanding of TTT dynamics. As additional understanding of phenomena driving TTTs is acquired, this type of MLP model can be used to populate unknowns (such as material impurity and other thermal mechanical properties) from past literature sources.

Physico-chemical properties of Moringa oleifera seed oil enzymatically interesterified with palm stearin and palm kernel oil and its potential application in food.

PubMed

Dollah, Sarafhana; Abdulkarim, Sabo Mohammed; Ahmad, Siti Hajar; Khoramnia, Anahita; Mohd Ghazali, Hasanah

2016-08-01

High oleic acid Moringa oleifera seed oil (MoO) has been rarely applied in food products due to the low melting point and lack of plasticity. Enzymatic interesterification (EIE) of MoO with palm stearin (PS) and palm kernel oil (PKO) could yield harder fat stocks that may impart desirable nutritional and physical properties. Blends of MoO and PS or PKO were examined for triacylglycerol (TAG) composition, thermal properties and solid fat content (SFC). EIE caused rearrangement of TAGs, reduction of U3 and increase of U2 S in MoO/PS blends while reduction of U3 and S3 following increase of S2 U and U2 S in MoO/PKO blends (U, unsaturated and S, saturated fatty acids). SFC measurements revealed a wide range of plasticity, enhancements of spreadability, mouthfeel and cooling effect for interesterified MoO/PS, indicating the possible application of these blends in margarines. However, interesterified MoO/PKO was not suitable in margarine application, while ice-cream may be formulated from these blends. A soft margarine formulated from MoO/PS 70:30 revealed high oxidative stability during 8 weeks storage with no significant changes in peroxide and p-anisidine values. EIE of fats with MoO allowed nutritional and oxidative stable plastic fats to be obtained, suitable for possible use in industrial food applications. © 2015 Society of Chemical Industry. © 2015 Society of Chemical Industry.

Grissom AFB, Indiana. Revised Uniform Summary of Surface Weather Observations (RUSSWO). Parts A-F.

DTIC Science & Technology

1983-07-01

8.8i 15.51 1b.3 27.1 7%39 USAFETAC 0 TOS N. OL. ), O~ Poyaw 0mo IMS rOR Omm V~ Jm- - - - - -- - - -- - - - - - - U A 2AL CLIMATOLOY B’ANCH 2 Z .Ac...2Z812.o3410.911 S. 945 ,.137 6.561 6.607 8.977 9.95612.15813.1I5 19.923 6~ 3 .46 927. 920 --U 9 R. Z 9Z5. 3 U. 9CC 93 Q 900 927. 1 C9 4. 13. 11.7 3j

Clinical usefulness of ursodeoxycholic acid for Japanese patients with autoimmune hepatitis

PubMed Central

Torisu, Yuichi; Nakano, Masanori; Takano, Keiko; Nakagawa, Ryo; Saeki, Chisato; Hokari, Atsushi; Ishikawa, Tomohisa; Saruta, Masayuki; Zeniya, Mikio

2017-01-01

AIM To evaluate the therapeutic effects of ursodeoxycholic acid (UDCA) on autoimmune hepatitis (AIH). METHODS A total 136 patients who were diagnosed with AIH were included in our study. All of the patients underwent a liver biopsy, and had at least a probable diagnosis on the basis of either the revised scoring system or the simplified scores. Initial treatment included UDCA monotherapy (Group U, n = 48) and prednisolone (PSL) monotherapy (Group P, n = 88). Group U was further classified into two subgroups according to the effect of UDCA: Patients who had achieved remission induction with UDCA monotherapy and showed no sign of relapse (Subgroup U1, n = 34) and patients who additionally received PSL during follow-up (Subgroup U2, n = 14). We compared the clinical and histological findings between each groups, and investigated factors contributing to the response to UDCA monotherapy. RESULTS In Group U, 34 patients (71%) achieved and maintained remission over 49 (range: 8-90) mo (Subgroup U1) and 14 patients (29%) additionally received PSL (Subgroup U2) during follow-up. Two patients in Subgroup U2 achieved remission induction once but additionally required PSL administration because of relapse (15 and 35 mo after the start of treatment). The remaining 12 patients in Subgroup U2 failed to achieve remission induction during follow-up, and PSL was added during 7 (range: 2-18) mo. Compared with Subgroup U2, Subgroup U1 had significantly lower alanine aminotransferase (ALT) levels at onset (124 IU/L vs 262 IU/L, P = 0.023) and a significantly higher proportion of patients with mild inflammation (A1) on histological examination (70.6% vs 35.7%, P = 0.025). When multivariate analysis was performed to identify factors contributing to the response to UDCA monotherapy, only a serum ALT level of 200 IU/L or lower was found to be associated with a significant difference (P = 0.013). CONCLUSION To prevent adverse events related to corticosteroids, UDCA monotherapy for AIH needs to be considered in patients with a serum ALT level of 200 IU/L or lower. PMID:28105259

15 CFR 744.7 - Restrictions on certain exports to and for the use of certain foreign vessels or aircraft.

Code of Federal Regulations, 2010 CFR

2010-01-01

..., other than fuel, is exported by U.S. airlines to their own aircraft abroad for their own use. (2) Exports to U.S. or Canadian Airline's Installation or Agent. Exports and reexports of the commodities described in paragraph (e) of this section, except fuel, may be made to a U.S. or Canadian airline's...

Atomization from a tantalum surface in graphite furnace atomic absorption spectrometry

NASA Astrophysics Data System (ADS)

Gregoire, D. C.; Chakrabarti, C. L.

The mechanism of atom formation of U, V, Mo, Ni, Mn, Cu and Mg atomized from pyrolytic graphite and tantalum metal surfaces has been studied. The mechanism of atom formation for U from a graphite tube atomizer is reported for the first time. The peak absorbance for U and Cu is increased by factors of 59.7 and 2.0, respectively, whereas that of V, Mo and Ni is reduced by several orders of magnitude when they are atomized from a tantalum metal surface. The peak absorbance of Mn and Mg is not appreciably affected by the material of the atomization surface. Interaction of Mn and Mg with the graphite surface and formation of their refractory carbides was found to be negligible. Uranium forms a refractory carbide when heated from a graphite surface.

SORPTION OF 2,3,7,8-TETRACHLORODIBENZO-P-DIOXIN TO SOILS FROM WATER/METHANOL MIXTURES

EPA Science Inventory

Sorption of ¹⁴C-labeled 2,3,7,8-tetrachlorodibenzo-p-dioxin (TCDD) to soils from water/methanol mixtures has been evaluated by batch shake testing. Uncontaminated soils from Times Beach, MO, were used in these experiments and ranged in fraction organic carbon (U...

The effect of thermomechanical processing on second phase particle redistribution in U-10 wt%Mo

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

Hu, Xiaohua; Wang, Xiaowo; Joshi, Vineet V.

2018-03-01

The multi-pass hot-rolling process of an annealed uranium-10 wt% molybdenum coupon was studied by plane-strain compression finite element modeling. Two point correlation function (2PCF) was used to analyze the carbide particle distribution after each rolling reduction. The hot rolling simulation results show that the alignment of UC particles along grain boundaries will rotate during rolling until it is parallel to the rolling direction, to form stringer-like distributions which are typically observed in rolled products that contain inclusions. 2PCF analysis of simulation shows that the interparticle spacing shrinks along the normal direction. The number of major peaks of 2PCF along NDmore » decreases after large reduction. The locations of major peaks indicate the inter-stringer distances.« less

Experimental study of the rearrangements of valence protons and neutrons amongst single-particle orbits during double- β decay in Mo 100

DOE PAGES

Freeman, S. J.; Sharp, D. K.; McAllister, S. A.; ...

2017-11-27

The rearrangements of protons and neutrons amongst the valence single-particle orbitals during double-beta decay of Mo-100 have been determined by measuring cross sections in (d, p), (p, d), (He-3, a), and (He-3, d) reactions on Mo-98,Mo-100 and Ru-100,Ru-102 targets. The deduced nucleon occupancies reveal significant discrepancies when compared with theoretical calculations; the same calculations have previously been used to determine the nuclear matrix element associated with the decay probability of double-beta decay of the Mo-100 system.

Experimental study of the rearrangements of valence protons and neutrons amongst single-particle orbits during double- β decay in Mo 100

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

Freeman, S. J.; Sharp, D. K.; McAllister, S. A.

The rearrangements of protons and neutrons amongst the valence single-particle orbitals during double-beta decay of Mo-100 have been determined by measuring cross sections in (d, p), (p, d), (He-3, a), and (He-3, d) reactions on Mo-98,Mo-100 and Ru-100,Ru-102 targets. The deduced nucleon occupancies reveal significant discrepancies when compared with theoretical calculations; the same calculations have previously been used to determine the nuclear matrix element associated with the decay probability of double-beta decay of the Mo-100 system.

High Power Diode Laser-Treated HP-HVOF and Twin Wire Arc-Sprayed Coatings for Fossil Fuel Power Plants

NASA Astrophysics Data System (ADS)

Mann, B. S.

2013-08-01

This article deals with high power diode laser (HPDL) surface modification of twin wire arc-sprayed (TWAS) and high pressure high velocity oxy-fuel (HP-HVOF) coatings to combat solid particle erosion occurring in fossil fuel power plants. To overcome solid particle impact wear above 673 K, Cr3C2-NiCr-, Cr3C2-CoNiCrAlY-, and WC-CrC-Ni-based HVOF coatings are used. WC-CoCr-based HVOF coatings are generally used below 673 K. Twin wire arc (TWA) spraying of Tafa 140 MXC and SHS 7170 cored wires is used for a wide range of applications for a temperature up to 1073 K. Laser surface modification of high chromium stainless steels for steam valve components and LPST blades is carried out regularly. TWA spraying using SHS 7170 cored wire, HP-HVOF coating using WC-CoCr powder, Ti6Al4V alloy, and high chromium stainless steels (X20Cr13, AISI 410, X10CrNiMoV1222, 13Cr4Ni, 17Cr4Ni) were selected in the present study. Using robotically controlled parameters, HPDL surface treatments of TWAS-coated high strength X10CrNiMoV1222 stainless steel and HP-HVOF-coated AISI 410 stainless steel samples were carried out and these were compared with HPDL-treated high chromium stainless steels and titanium alloy for high energy particle impact wear (HEPIW) resistance. The HPDL surface treatment of the coatings has improved the HEPIW resistance manifold. The improvement in HPDL-treated stainless steels and titanium alloys is marginal and it is not comparable with that of HPDL-treated coatings. These coatings were also compared with "as-sprayed" coatings for fracture toughness, microhardness, microstructure, and phase analyses. The HEPIW resistance has a strong relationship with the product of fracture toughness and microhardness of the HPDL-treated HP-HVOF and TWAS SHS 7170 coatings. This development opens up a possibility of using HPDL surface treatments in specialized areas where the problem of HEPIW is very severe. The HEPIW resistance of HPDL-treated high chromium stainless steels and titanium alloys, HPDL-treated TWAS SHS 7170 and HP-HVOF coatings, and their micrographs and X-ray diffraction analysis is reported in this article.

Monitoring the growth of polyoxomolybdate nanoparticles in suspension by flow field-flow fractionation.

PubMed

Chen, Bailin; Jiang, Huijian; Zhu, Yan; Cammers, Arthur; Selegue, John P

2005-03-30

We follow the evolution of polyoxomolybdate nanoparticles in suspensions derived from the keplerate (NH4)42[MoVI72MoV60O372(CH3CO2)30(H2O)72].ca..300H2O.ca..10CH3CO2NH4 ({Mo132}) by flow field-flow fractionation (FlFFF) to monitor the particle-size distribution in situ, atomic force and high-resolution transmission electron microscopy (AFM, SEM, and HRTEM) to confirm particle sizes, inductively coupled plasma-optical emission spectrometry (ICP-OES) to determine the Mo content of the FlFFF-separated fractions, and UV/visible spectroscopy to confirm the identity of the species in suspension. We observe the formation of 3-75-nm polyoxomolybdate particles in suspension and the dynamic growth of {Mo132} crystals.

Simulation of the MoEDAL experiment

NASA Astrophysics Data System (ADS)

King, Matthew; MoEDAL Collaboration

2016-04-01

The MoEDAL experiment (Monopole and Exotics Detector at the LHC) is designed to directly search for magnetic monopoles and other highly ionising stable or meta-stable particles at the LHC. The MoEDAL detector comprises an array of plastic track detectors and aluminium trapping volumes around the P8 intersection region, opposite from the LHCb detector. TimePix devices are also installed for monitoring of the experiment. As MoEDAL mostly employs passive detectors the software development focusses on particle simulation, rather than digitisation or reconstruction. Here, we present the current status of the MoEDAL simulation software. Specifically, the development of a material description of the detector and simulations of monopole production and propagation at MoEDAL.

CoCrMo alloy vs. UHMWPE Particulate Implant Debris Induces Sex Dependent Aseptic Osteolysis Responses In Vivo using a Murine Model

PubMed Central

Landgraeber, Stefan; Samelko, Lauryn; McAllister, Kyron; Putz, Sebastian; Jacobs, Joshua.J.; Hallab, Nadim James

2018-01-01

Background: The rate of revision for some designs of total hip replacements due to idiopathic aseptic loosening has been reported as higher for women. However, whether this is environmental or inherently sex-related is not clear. Objective: Can particle induced osteolysis be sex dependent? And if so, is this dependent on the type of implant debris (e.g. metal vs polymer)? The objective of this study was to test for material dependent inflammatory osteolysis that may be linked to sex using CoCrMo and implant grade conventional polyethylene (UHMWPE), using an in vivo murine calvaria model. Methods: Healthy 12 week old female and male C57BL/6J mice were treated with UHMWPE (1.0um ECD) or CoCrMo particles (0.9um ECD) or received sham surgery. Bone resorption was assessed by micro-computed tomography, histology and histomorphometry on day 12 post challenge. Results: Female mice that received CoCrMo particles showed significantly more inflammatory osteolysis and bone destruction compared to the females who received UHMWPE implant debris. Moreover, females challenged with CoCrMo particles exhibited 120% more inflammatory bone loss compared to males (p<0.01) challenged with CoCrMo implant debris (but this was not the case for UHMWPE particles). Conclusion: We demonstrated sex-specific differences in the amount of osteolysis resulting from CoCrMo particle challenge. This suggests osteo-immune responses to metal debris are preferentially higher in female compared to male mice, and supports the contention that there may be inherent sex related susceptibility to some types of implant debris. PMID:29785221

Performance evaluation of platinum-molybdenum carbide nanocatalysts with ultralow platinum loading on anode and cathode catalyst layers of proton exchange membrane fuel cells

NASA Astrophysics Data System (ADS)

Saha, Shibely; Cabrera Rodas, José Andrés; Tan, Shuai; Li, Dongmei

2018-02-01

An alternative catalyst platform, consisting of a phase-pure transition carbide (TMC) support and Pt nanoparticles (NPs) in the range of subnanometer to < 2.7 nm, is established that can be used in both anode and cathode catalyst layers. While some TMCs with low Pt loadings have demonstrated similar activity as commercial Pt catalyst in idealized disk electrode screening tests, few to none have been applied in a realistic fuel cell membrane electrode assembly (MEA). We recently reported that β-Mo2C hollow nanotubes modified with Pt NPs via atomic layer deposition (ALD) possess better activity and durability than 20% Pt/C. This paper presents systematic evaluation of the Pt/Mo2C catalysts in a MEA, investigating effects of different MEA preparation techniques, gas diffusion layers (GDL) and various Pt loadings in the ultralow range (<0.04 mg/cm2) on MEA performance. Most importantly, we demonstrate, for the first time, that Pt/Mo2C catalyst on both anode and cathode, with a loading of 0.02 mg (Pt) cm-2, generated peak power density of 414 mW cm-2 that corresponds to 10.35 kWgPt-1 using hydrogen (H2) and oxygen (O2). Accelerated degradation tests (ADT) on Pt/Mo2C catalysts show 111% higher power density than commercial 20% Pt/C after the vigorous ADT.

National Dam Safety Program. Bowling Green Dam, (MO 10262) Mississippi - Kaskaskia - St. Louis Basin, Pike County, Missouri. Phase I Inspection Report.

DTIC Science & Technology

1978-12-01

4 L-7 hiI<±b fd .rvc~ a .3",d,! i:; ba ’-ii~ y :nztructl~n :).L a -- flattctod r~-%3 i Pazo u-3 and r.). 1yzj. f r 3n- the road-rmy, S11eet 13...Program " Final Report Bowling Green Reservoir Dam (MO 10262) 6. PERFORMING-ORG. REPORT NUMBER Pike County, Missouri 7. AUTHOR( a ) S. CONTRACT OR GRANT...if possible. If a classification is required, identify the classified items on the page by the appropriate symbol. CC; -ETION GUIDE General. Make

First-principles study on influence of molybdenum on acicular ferrite formation on TiC particles in microallyed steels

NASA Astrophysics Data System (ADS)

Hua, Guomin; Li, Changsheng; Cheng, Xiaonong; Zhao, Xinluo; Feng, Quan; Li, Zhijie; Li, Dongyang; Szpunar, Jerzy A.

2018-01-01

In this study, influences of molybdenum on acicular ferrite formation on precipitated TiC particles are investigated from thermodynamic and kinetic respects. In thermodynamics, Segregation of Mo towards austenite/TiC interface releases the interfacial energy and induces phase transformation from austenite to acicular ferrite on the precipitated TiC particles. The Phase transformation can be achieved by displacive deformation along uniaxial Bain path. In addition, the segregation of Mo atom will also lead to the enhanced stability of ferrite in comparison with austenite no matter at low temperature or at high temperature. In kinetics, the Mo solute in acicular ferrite can effectively suppress the diffusion of carbon atoms, which ensures that orientation relationship between acicular ferrite and austenitized matrix can be satisfied during the diffusionless phase transformation. In contrast to ineffectiveness of TiC particles, the alloying Mo element can facilitate the formation of acicular ferrite on precipitated TiC particles, which is attributed to the above thermodynamic and kinetic reasons. Furthermore, Interfacial toughness and ductility of as-formed acicular ferrite/TiC interface can be improved simultaneously by segregation of Mo atom.

The effects of temperature and exercise training on swimming performance in juvenile qingbo (Spinibarbus sinensis).

PubMed

Pang, Xu; Yuan, Xing-Zhong; Cao, Zhen-Dong; Fu, Shi-Jian

2013-01-01

To investigate the effects of temperature and exercise training on swimming performance in juvenile qingbo (Spinibarbus sinensis), we measured the following: (1) the resting oxygen consumption rate (MO(2rest)), critical swimming speed (U(crit)) and active oxygen consumption rate (MO(2active)) of fish at acclimation temperatures of 10, 15, 20, 25 and 30 °C and (2) the MO(2rest), U(crit) and MO(2active) of both exercise-trained (exhaustive chasing training for 14 days) and control fish at both low and high acclimation temperatures (15 and 25 °C). The relationship between U(crit) and temperature (T) approximately followed a bell-shaped curve as temperature increased: U(crit) = 8.21/{1 + [(T - 27.2)/17.0]²} (R² = 0.915, P < 0.001, N = 40). The optimal temperature for maximal U(crit) (8.21 BL s(-1)) in juvenile qingbo was 27.2 °C. Both the MO(2active) and the metabolic scope (MS, MO(2active) - MO(2rest)) of qingbo increased with temperature from 10 to 25 °C (P < 0.05), but there were no significant differences between fish acclimated to 25 and 30 °C. The relationships between MO(2active) or MS and temperature were described as MO(2active) = 1,214.29 /{1 + [(T - 28.8)/10.6]²} (R² = 0.911, P < 0.001, N = 40) and MS = 972.67/{1 + [(T - 28.0)/9.34]²} (R² = 0.878, P < 0.001, N = 40). The optimal temperatures for MO(2active) and MS in juvenile qingbo were 28.8 and 28.0 °C, respectively. Exercise training resulted in significant increases in both U(crit) and MO(2active) at a low temperature (P < 0.05), but training exhibited no significant effect on either U(crit) or MO(2active) at a high temperature. These results suggest that exercise training had different effects on swimming performance at different temperatures. These differences may be related to changes in aerobic metabolic capability, arterial oxygen delivery, available dissolved oxygen, imbalances in ion fluxes and stimuli to remodel tissues with changes in temperature.

Characterization of Delayed-Particle Emission Signatures for Pyroprocessing. Part 1: ABTR Fuel Assembly.

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

Durkee, Jr., Joe W.

A three-part study is conducted using the MCNP6 Monte Carlo radiation-transport code to calculate delayed-neutron (DN) and delayed-gamma (DG) emission signatures for nondestructive assay (NDA) metal-fuel pyroprocessing. In Part 1, MCNP6 is used to produce irradiation-induced used nuclear fuel (UNF) isotopic inventories for an Argonne National Laboratory (ANL) Advanced Burner Test Reactor (ABTR) preconceptual design fuel assembly (FA) model. The initial fuel inventory consists of uranium mixed with light-water-reactor transuranic (TRU) waste and 10 wt% zirconium (U-LWR-SFTRU-10%Zr). To facilitate understanding, parametric evaluation is done using models for 3% and 5% initial 235U a% enrichments, burnups of 5, 10, 15, 20,more » 30, …, 120 GWd/MTIHM, and 3-, 5-, 10-, 20-, and 30- year cooling times. Detailed delayed-particle radioisotope source terms for the irradiate FA are created using BAMF-DRT and SOURCES3A. Using simulation tallies, DG activity ratios (DGARs) are developed for 134Cs/ 137Cs 134Cs/ 154Eu, and 154Eu/ 137Cs markers as a function of (1) burnup and (2) actinide mass, including elemental uranium, neptunium, plutonium, americium, and curium. Spectral-integrated DN emission is also tallied. The study reveals a rich assortment of DGAR behavior as a function of DGAR type, enrichment, burnup, and cooling time. Similarly, DN emission plots show variation as a function of burnup and of actinide mass. Sensitivity of DGAR and DN signatures to initial 235U enrichment, burnup, and cooling time is evident. Comparisons of the ABTR radiation signatures and radiation signatures previously reported for a generic Westinghouse oxide-fuel assembly indicate that there are pronounced differences in the ABTR and Westinghouse oxide-fuel DN and DG signatures. These differences are largely attributable to the initial TRU inventory in the ABTR fuel. The actinide and nonactinide inventories for the FA models serve as source materials for the pre- and postelectrorefining models to be reported in Parts 2 and 3.« less

Development of Plasma-Sprayed Molybdenum Carbide-Based Anode Layers with Various Metal Oxides for SOFC

NASA Astrophysics Data System (ADS)

Faisal, N. H.; Ahmed, R.; Katikaneni, S. P.; Souentie, S.; Goosen, M. F. A.

2015-12-01

Air plasma-sprayed (APS) coatings provide an ability to deposit a range of novel fuel cell materials at competitive costs. This work develops three separate types of composite anodes (Mo-Mo2C/Al2O3, Mo-Mo2C/ZrO2, Mo-Mo2C/TiO2) using a combination of APS process parameters on Hastelloy®X for application in intermediate temperature proton-conducting solid oxide fuel cells. Commercially available carbide of molybdenum powder catalyst (Mo-Mo2C) and three metal oxides (Al2O3, ZrO2, TiO2) was used to prepare three separate composite feedstock powders to fabricate three different anodes. Each of the modified composition anode feedstock powders included a stoichiometric weight ratio of 0.8:0.2. The coatings were characterized by scanning electron microscopy, energy dispersive spectroscopy, x-ray diffraction, nanoindentation, and conductivity. We report herein that three optimized anode layers of thicknesses between 200 and 300 µm and porosity as high as 20% for Mo-Mo2C/Al2O3 (250-µm thick) and Mo-Mo2C/TiO2 (300 µm thick) and 17% for Mo-Mo2C/ZrO2 (220-µm thick), controllable by a selection of the APS process parameters with no addition of sacrificial pore-forming material. The nanohardness results indicate the upper layers of the coatings have higher values than the subsurface layers in coatings with some effect of the deposition on the substrate. Mo-Mo2C/ZrO2 shows high electrical conductivity.

The effects of particle swarm optimization algorithm on volume ignition gain of Proton-Lithium (7) pellets

NASA Astrophysics Data System (ADS)

Livari, As. Ali; Malekynia, B.; Livari, Ak. A.; Khoda-Bakhsh, R.

2017-11-01

When it was found out that the ignition of nuclear fusion hinges upon input energy laser, the efforts in order to make giant lasers began, and energy gains of DT fuel were obtained. But due to the neutrons generation and emitted radioactivity from DT reaction, gains of fuels like Proton-Lithium (7) were also adverted. Therefore, making larger and powerful lasers was followed. Here, using new versions of particle swarm optimization algorithm, it will be shown that available maximum gain of Proton-Lithium (7) is reached only at energies about 1014 J, and not only the highest input energy is not helpful but the efficiency is also decreased.

Development of multilayer perceptron networks for isothermal time temperature transformation prediction of U-Mo-X alloys

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

Johns, Jesse M.; Burkes, Douglas

In this work, a multilayered perceptron (MLP) network is used to develop predictive isothermal time-temperature-transformation (TTT) models covering a range of U-Mo binary and ternary alloys. The selected ternary alloys for model development are U-Mo-Ru, U-Mo-Nb, U-Mo-Zr, U-Mo-Cr, and U-Mo-Re. These model’s ability to predict 'novel' U-Mo alloys is shown quite well despite the discrepancies between literature sources for similar alloys which likely arise from different thermal-mechanical processing conditions. These models are developed with the primary purpose of informing experimental decisions. Additional experimental insight is necessary in order to reduce the number of experiments required to isolate ideal alloys. Thesemore » models allow test planners to evaluate areas of experimental interest; once initial tests are conducted, the model can be updated and further improve follow-on testing decisions. The model also improves analysis capabilities by reducing the number of data points necessary from any particular test. For example, if one or two isotherms are measured during a test, the model can construct the rest of the TTT curve over a wide range of temperature and time. This modeling capability reduces the cost of experiments while also improving the value of the results from the tests. The reduced costs could result in improved material characterization and therefore improved fundamental understanding of TTT dynamics. As additional understanding of phenomena driving TTTs is acquired, this type of MLP model can be used to populate unknowns (such as material impurity and other thermal mechanical properties) from past literature sources.« less

Genotoxic potential of diesel exhaust particles from the combustion of first- and second-generation biodiesel fuels-the FuelHealth project.

PubMed

Kowalska, Magdalena; Wegierek-Ciuk, Aneta; Brzoska, Kamil; Wojewodzka, Maria; Meczynska-Wielgosz, Sylwia; Gromadzka-Ostrowska, Joanna; Mruk, Remigiusz; Øvrevik, Johan; Kruszewski, Marcin; Lankoff, Anna

2017-11-01

Epidemiological data indicate that exposure to diesel exhaust particles (DEPs) from traffic emissions is associated with higher risk of morbidity and mortality related to cardiovascular and pulmonary diseases, accelerated progression of atherosclerotic plaques, and possible lung cancer. While the impact of DEPs from combustion of fossil diesel fuel on human health has been extensively studied, current knowledge of DEPs from combustion of biofuels provides limited and inconsistent information about its mutagenicity and genotoxicity, as well as possible adverse health risks. The objective of the present work was to compare the genotoxicity of DEPs from combustion of two first-generation fuels, 7% fatty acid methyl esters (FAME) (B7) and 20% FAME (B20), and a second-generation 20% FAME/hydrotreated vegetable oil (SHB: synthetic hydrocarbon biofuel) fuel. Our results revealed that particulate engine emissions from each type of biodiesel fuel induced genotoxic effects in BEAS-2B and A549 cells, manifested as the increased levels of single-strand breaks, the increased frequencies of micronuclei, or the deregulated expression of genes involved in DNA damage signaling pathways. We also found that none of the tested DEPs showed the induction of oxidative DNA damage and the gamma-H2AX-detectable double-strand breaks. The most pronounced differences concerning the tested particles were observed for the induction of single-strand breaks, with the greatest genotoxicity being associated with the B7-derived DEPs. The differences in other effects between DEPs from the different biodiesel blend percentage and biodiesel feedstock were also observed, but the magnitude of these variations was limited.

Study of the dynamics of the MoO2-Mo2C system for catalytic partial oxidation reactions

NASA Astrophysics Data System (ADS)

Cuba Torres, Christian Martin

On a global scale, the energy demand is largely supplied by the combustion of non-renewable fossil fuels. However, their rapid depletion coupled with environmental and sustainability concerns are the main drivers to seek for alternative energetic strategies. To this end, the sustainable generation of hydrogen from renewable resources such as biodiesel would represent an attractive alternative solution to fossil fuels. Furthermore, hydrogen's lower environmental impact and greater independence from foreign control make it a strong contender for solving this global problem. Among a wide variety of methods for hydrogen production, the catalytic partial oxidation offers numerous advantages for compact and mobile fuel processing systems. For this reaction, the present work explores the versatility of the Mo--O--C catalytic system under different synthesis methods and reforming conditions using methyl oleate as a surrogate biodiesel. MoO2 exhibits good catalytic activity and exhibits high coke-resistance even under reforming conditions where long-chain oxygenated compounds are prone to form coke. Moreover, the lattice oxygen present in MoO2 promotes the Mars-Van Krevelen mechanism. Also, it is introduced a novel beta-Mo2C synthesis by the in-situ formation method that does not utilize external H2 inputs. Herein, the MoO 2/Mo2C system maintains high catalytic activity for partial oxidation while the lattice oxygen serves as a carbon buffer for preventing coke formation. This unique feature allows for longer operation reforming times despite slightly lower catalytic activity compared to the catalysts prepared by the traditional temperature-programmed reaction method. Moreover, it is demonstrated by a pulse reaction technique that during the phase transformation of MoO2 to beta-Mo2C, the formation of Mo metal as an intermediate is not responsible for the sintering of the material wrongly assumed by the temperature-programmed method.

U.S. Light-duty Vehicle Air Conditioning Fuel Use and the Impact of Four Solar/Thermal Control Technologies

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

Rugh, John P; Kekelia, Bidzina; Kreutzer, Cory J

The U.S. uses 7.6 billion gallons of fuel per year for vehicle air conditioning (A/C), equivalent to 5.7 percent of the total national light-duty vehicle (LDV) fuel use. This equates to 30 gallons/year per vehicle, or 23.5 grams (g) of carbon dioxide (CO2) per mile, for an average U.S. vehicle. A/C is a significant contribution to national fuel use; therefore, technologies that reduce A/C loads may reduce operational costs, A/C fuel use, and CO2 emissions. Since A/C is not operated during standard EPA fuel economy testing protocols, EPA provides off-cycle credits to encourage OEMs to implement advanced A/C technologies thatmore » reduce fuel use in the real world. NREL researchers assessed thermal/solar off-cycle credits available in the U.S. Environmental Protection Agency's (EPA's) Final Rule for Model Year 2017 and Later Light-Duty Vehicle Greenhouse Gas Emissions and Corporate Average Fuel Economy. Credits include glazings, solar reflective paint, and passive and active cabin ventilation. Implementing solar control glass reduced CO2 emissions by 2.0 g/mi, and solar reflective paint resulted in a reduction of 0.8 g/mi. Active and passive ventilation strategies only reduced emissions by 0.1 and 0.2 g/mi, respectively. The national-level analysis process is powerful and general; it can be used to determine the impact of a wide range of new vehicle thermal technologies on fuel use, EV range, and CO2 emissions.« less

Reductions in aircraft particulate emissions due to the use of Fischer-Tropsch fuels

NASA Astrophysics Data System (ADS)

Beyersdorf, A. J.; Timko, M. T.; Ziemba, L. D.; Bulzan, D.; Corporan, E.; Herndon, S. C.; Howard, R.; Miake-Lye, R.; Thornhill, K. L.; Winstead, E.; Wey, C.; Yu, Z.; Anderson, B. E.

2014-01-01

The use of alternative fuels for aviation is likely to increase due to concerns over fuel security, price stability, and the sustainability of fuel sources. Concurrent reductions in particulate emissions from these alternative fuels are expected because of changes in fuel composition including reduced sulfur and aromatic content. The NASA Alternative Aviation Fuel Experiment (AAFEX) was conducted in January-February 2009 to investigate the effects of synthetic fuels on gas-phase and particulate emissions. Standard petroleum JP-8 fuel, pure synthetic fuels produced from natural gas and coal feedstocks using the Fischer-Tropsch (FT) process, and 50% blends of both fuels were tested in the CFM-56 engines on a DC-8 aircraft. To examine plume chemistry and particle evolution with time, samples were drawn from inlet probes positioned 1, 30, and 145 m downstream of the aircraft engines. No significant alteration to engine performance was measured when burning the alternative fuels. However, leaks in the aircraft fuel system were detected when operated with the pure FT fuels as a result of the absence of aromatic compounds in the fuel. Dramatic reductions in soot emissions were measured for both the pure FT fuels (reductions in mass of 86% averaged over all powers) and blended fuels (66%) relative to the JP-8 baseline with the largest reductions at idle conditions. At 7% power, this corresponds to a reduction from 7.6 mg kg-1 for JP-8 to 1.2 mg kg-1 for the natural gas FT fuel. At full power, soot emissions were reduced from 103 to 24 mg kg-1 (JP-8 and natural gas FT, respectively). The alternative fuels also produced smaller soot (e.g., at 85% power, volume mean diameters were reduced from 78 nm for JP-8 to 51 nm for the natural gas FT fuel), which may reduce their ability to act as cloud condensation nuclei (CCN). The reductions in particulate emissions are expected for all alternative fuels with similar reductions in fuel sulfur and aromatic content regardless of the feedstock. As the plume cools downwind of the engine, nucleation-mode aerosols form. For the pure FT fuels, reductions (94% averaged over all powers) in downwind particle number emissions were similar to those measured at the exhaust plane (84%). However, the blended fuels had less of a reduction (reductions of 30-44%) than initially measured (64%). The likely explanation is that the reduced soot emissions in the blended fuel exhaust plume results in promotion of new particle formation microphysics, rather than coating on pre-existing soot particles, which is dominant in the JP-8 exhaust plume. Downwind particle volume emissions were reduced for both the pure (79 and 86% reductions) and blended FT fuels (36 and 46%) due to the large reductions in soot emissions. In addition, the alternative fuels had reduced particulate sulfate production (near zero for FT fuels) due to decreased fuel sulfur content. To study the formation of volatile aerosols (defined as any aerosol formed as the plume ages) in more detail, tests were performed at varying ambient temperatures (-4 to 20 °C). At idle, particle number and volume emissions were reduced linearly with increasing ambient temperature, with best fit slopes corresponding to -8 × 1014 particles (kg fuel)-1 °C-1 for particle number emissions and -10 mm3 (kg fuel)-1 °C-1 for particle volume emissions. The temperature dependency of aerosol formation can have large effects on local air quality surrounding airports in cold regions. Aircraft-produced aerosols in these regions will be much larger than levels expected based solely on measurements made directly at the engine exit plane. The majority (90% at idle) of the volatile aerosol mass formed as nucleation-mode aerosols, with a smaller fraction as a soot coating. Conversion efficiencies of up to 2.8% were measured for the partitioning of gas-phase precursors (unburned hydrocarbons and SO2) to form volatile aerosols. Highest conversion efficiencies were measured at 45% power.

Manufacturing Methods and Technology. Project Execution Report.

DTIC Science & Technology

1984-05-01

U5/62"/3. 20 "°"’°’. . .. - 2 °* ’%7’. 471 FINAL STATUS REPORTS RLCEIVtD vURINU 2Nv HALF, CY6- . ( LLNT I NUE O) - o aD ’ 281 LUNSLRVATiUi4 bF tNERGY AT...KO IA w. hJO 0. xV W-2 -C -a aZ. 0 0 0 m C9 - I 0 MO m4a. W 19 6u Z ) -- 03. zWI 2 M 3 U- P- OCw 43 3B *u. 00 CKa I.-hAV . 40 2- w cc . U6 U- *~- -02

Phase stability and electronic structure of UMo2Al20: A first-principles study

NASA Astrophysics Data System (ADS)

Liu, Peng-Chuang; Xian, Ya-Jiang; Wang, Xin; Zhang, Yu-Ting; Zhang, Peng-Cheng

2017-09-01

In this paper, the phase stability of UMo2Al20 was explored using cluster formula in combination with first-principles calculations. Cluster formula analysis uncovered that the compound was composed of two principal clusters, i.e. [Mo-Al12] and [U-Al16]. The electronic interactions between U, Mo and Al atoms in this compound were discussed using elastic property, Bader charges and energy-resolved local bonding analysis, as well as the electronic interactions between Mo and Al atoms in [Mo-Al12] cluster and between U and Al atoms in [U-Al16] cluster. It revealed that UMo2Al20 satisfied the mechanical stability criterion for cubic system, and exhibited near ionic bonding character with weak bonding directionality. The calculations within both standard DFT and HSE frameworks demonstrated that U and Al atoms acted as an electron donor while Mo atoms acted as electron acceptor. The intrinsic stability of UMo2Al20 mainly stemmed from the bonding states of Mo-Al bonds and Al-Al bonds in [Mo-Al12] cluster. These calculations provide a further insight on the CeCr2Al20-type ternary compounds.

Tumorigenic Evaluation of Jet Fuels JP-TS and JP-7.

DTIC Science & Technology

1991-04-01

DTIC AL-TR-1991 0020 3 ELECTE0 AD-A252 012 JUN 2 6 1992• • TUMORIGENIC EVALUATION OF JET FUELS JP-TS AND JP-7 E. R. Kinkead C. L. Gaworski C. D...Evaluation of Jet Fuels JP-TS and JP-7. The research described in this report began in March 1981 and was completed in February 1991 under U.S. Air Force...of jet engines in military and commercial aircraft has led to the development of a number of petroleum distillate fuels with special properties. These

Effect of Heat Treatment on the Microstructure and Hardness of 17Cr-0.17N-0.43C-1.7 Mo Martensitic Stainless Steel

NASA Astrophysics Data System (ADS)

Krishna, S. Chenna; Gangwar, Narendra Kumar; Jha, Abhay K.; Pant, Bhanu; George, Koshy M.

2015-04-01

The microstructure and hardness of a nitrogen-containing martensitic stainless steel were investigated as a function of heat treatment using optical microscopy, electron microscopy, amount of retained austenite, and hardness measurement. The steel was subjected to three heat treatments: hardening, cryo treatment, and tempering. The hardness of the steel in different heat-treated conditions ranged within 446-620 HV. The constituents of microstructure in hardened condition were lath martensite, retained austenite, M23C6, M7C3, MC carbides, and M(C,N) carbonitrides. Upon tempering at 500 °C, two new phases have precipitated: fine spherical Mo2C carbides and needle-shaped Cr2N particles.

Dual-Doped Molybdenum Trioxide Nanowires: A Bifunctional Anode for Fiber-Shaped Asymmetric Supercapacitors and Microbial Fuel Cells.

PubMed

Yu, Minghao; Cheng, Xinyu; Zeng, Yinxiang; Wang, Zilong; Tong, Yexiang; Lu, Xihong; Yang, Shihe

2016-06-01

A novel in situ N and low-valence-state Mo dual doping strategy was employed to significantly improve the conductivity, active-site accessibility, and electrochemical stability of MoO3 , drastically boosting its electrochemical properties. Consequently, our optimized N-MoO3-x nanowires exhibited exceptional performances as a bifunctional anode material for both fiber-shaped asymmetric supercapacitors (ASCs) and microbial fuel cells (MFCs). The flexible fiber-shaped ASC and MFC device based on the N-MoO3-x anode could deliver an unprecedentedly high energy density of 2.29 mWh cm(-3) and a remarkable power density of 0.76 μW cm(-1) , respectively. Such a bifunctional fiber-shaped N-MoO3-x electrode opens the way to integrate the electricity generation and storage for self-powered sources. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

U-10Mo Sample Preparation and Examination using Optical and Scanning Electron Microscopy

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

Prabhakaran, Ramprashad; Joshi, Vineet V.; Rhodes, Mark A.

2016-10-01

The purpose of this document is to provide guidelines to prepare specimens of uranium alloyed with 10 weight percent molybdenum (U-10Mo) for optical metallography and scanning electron microscopy. This document also provides instructions to set up an optical microscope and a scanning electron microscope to analyze U-10Mo specimens and to obtain the required information.

U-10Mo Sample Preparation and Examination using Optical and Scanning Electron Microscopy

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

Prabhakaran, Ramprashad; Joshi, Vineet V.; Rhodes, Mark A.

2016-03-30

The purpose of this document is to provide guidelines to prepare specimens of uranium alloyed with 10 weight percent molybdenum (U-10Mo) for optical metallography and scanning electron microscopy. This document also provides instructions to set up an optical microscope and a scanning electron microscope to analyze U-10Mo specimens and to obtain the required information.

Catalytic conversion of isophorone to jet-fuel range aromatic hydrocarbons over a MoO(x)/SiO2 catalyst.

PubMed

Chen, Fang; Li, Ning; Wang, Wentao; Wang, Aiqin; Cong, Yu; Wang, Xiaodong; Zhang, Tao

2015-07-28

For the first time, jet fuel range C8-C9 aromatic hydrocarbons were synthesized in high carbon yield (∼80%) by the catalytic conversion of isophorone over MoO(x)/SiO2 at atmospheric pressure. A possible reaction pathway was proposed according to the control experiments and the intermediates generated during the reaction.

Technology Validation: Fuel Cell Bus Evaluations

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

Eudy, Leslie

This presentation describing the FY 2016 accomplishments for the National Renewable Energy Laboratory's Fuel Cell Bus Evaluations project was presented at the U.S. Department of Energy Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Meeting, June 7, 2016.

Topical Hazard Evaluation Program of Candidate Insect Repellent AI3- 39053a

DTIC Science & Technology

1987-05-01

StudyNo. 75-51-O 71-R7 Mnay 19R71 12 PERSONAL AUTHOR(S) William T. Huebsam, SGT, U.S.A.; Maurice H. Weeks 1I8. TYPE Of REPORT 13b TIME COVERED 114...ARMY u.S. AUT ENVIVIf[NNIAL HYGINE AMIdT ASCROELN PROVING GROUND. MARYLAND M1蚉 HSHB-MO-T TOPICAL HAZARD EVALUATION PROGRAM OF CANDIDATE INSECT

PARFUME Theory and Model basis Report

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

Darrell L. Knudson; Gregory K Miller; G.K. Miller

2009-09-01

The success of gas reactors depends upon the safety and quality of the coated particle fuel. The fuel performance modeling code PARFUME simulates the mechanical, thermal and physico-chemical behavior of fuel particles during irradiation. This report documents the theory and material properties behind vari¬ous capabilities of the code, which include: 1) various options for calculating CO production and fission product gas release, 2) an analytical solution for stresses in the coating layers that accounts for irradiation-induced creep and swelling of the pyrocarbon layers, 3) a thermal model that calculates a time-dependent temperature profile through a pebble bed sphere or amore » prismatic block core, as well as through the layers of each analyzed particle, 4) simulation of multi-dimensional particle behavior associated with cracking in the IPyC layer, partial debonding of the IPyC from the SiC, particle asphericity, and kernel migration (or amoeba effect), 5) two independent methods for determining particle failure probabilities, 6) a model for calculating release-to-birth (R/B) ratios of gaseous fission products that accounts for particle failures and uranium contamination in the fuel matrix, and 7) the evaluation of an accident condition, where a particle experiences a sudden change in temperature following a period of normal irradiation. The accident condi¬tion entails diffusion of fission products through the particle coating layers and through the fuel matrix to the coolant boundary. This document represents the initial version of the PARFUME Theory and Model Basis Report. More detailed descriptions will be provided in future revisions.« less

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.

Acceptance Test Data for Candidate AGR-5/6/7 TRISO Particle Batches BWXT Coater Batches 93165 93172 Defective IPyC Fraction and Pyrocarbon Anisotropy

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

Helmreich, Grant W.; Hunn, John D.; Skitt, Darren J.

2017-03-01

Coated particle fuel batches J52O-16-93165, 93166, 93168, 93169, 93170, and 93172 were produced by Babcock and Wilcox Technologies (BWXT) for possible selection as fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR). Some of these batches may alternately be used as demonstration coated particle fuel for other experiments. Each batch was coated in a 150-mm-diameter production-scale fluidized-bed chemical vapor deposition (CVD) furnace. Tristructural isotropic (TRISO) coatings were deposited on 425-μm-nominal-diameter spherical kernels from BWXT lot J52R-16-69317 containing a mixture of 15.5%-enriched uranium carbide andmore » uranium oxide (UCO). The TRISO coatings consisted of four consecutive CVD layers: a ~50% dense carbon buffer layer with 100-μm-nominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μmnominal thickness. The TRISO-coated particle batches were sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batches were designated by appending the letter A to the end of the batch number (e.g., 93165A).« less

Phase composition and fine structure of 0.18C-1Cr-3Ni-1Mo-Fe steel after plasma-electrolytic treatment

NASA Astrophysics Data System (ADS)

Popova, Natalya; Bayatanova, Lyayla; Nikonenko, Elena; Skakov, Mazhyn; Kozlov, Eduard

2017-01-01

The paper presents the transmission electron microscopy (TEM) investigation of 0.18C-1Cr-3Mn-1Mo- Fe steel specimens to study phase transitions and modification of fine structure after plasma-electrolytic treatment (carbonitriding at 850°C during 5 min). TEM investigations involve two points: on the specimen surface and at ˜40 µm distance from it. The experiments show that the structure in the original state is a mixture consisting of ferrite and perlite grains. Carbonitriding results in a considerable modification of the quality and quantity of steel structure. Thus, on the surface, α-phase is represented by lamellar martensite, while at ˜40 µm depth - by massive and lamellar martensite tempered at low and high temperatures. Moreover, on the subsurface of the martensite plates' boundaries retained austenite layers are observed, while inside plates the particles of alloyed cementite, carbonitrides of M23(C,N)6, M2C0.61N0.39, M6,2C3,5N0,3, M(C,N)2, Cr12Fe32Mo7Ni7 types, and β-graphite are present. In the specimen at the depth of ˜40 µm, retained austenite layers are observed on the boundaries of martensite laths and plates, while inside plates only the particles of alloyed cementite and M23(C,N)6 carbonitride are formed.

DoD Statistical Report on the Military Retirement System. Fiscal Year 1983.

DTIC Science & Technology

1983-01-01

Qne 00000a~a cocoa 00000 0=00m obot 0=o- 00000 00000 coa=cc. 0 a w a .4 0aa coo -- a- aCca C.j a. a ... 0000 a-=.. 00000 00000 00000 00000 00000...4rn , c xV Mo NC.~ Mnoio Or m -.Y c 0.K m. f 0;: 1a -a61 W. 42 *K2% ffM W 0" _F Ww~.. 0K#-: =MWVNC 6O𔃺 NK I0 cc~ ID I 0 ,W NK :IW𔃺 C,*ONg u0:o-mo W...CC# N NIN N N ##CC#CCCC4V)4MMM . or zzzrjj-jjJ.Jjj-J 2 N U0 - l l 7C= ffM M 0’-w 2 N4EfW ’ 019.03,0 1 IM -n INIi LrW U,0 lffO6w lbD K 227 ~ 222 .J

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

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

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

2012-10-01

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

Effect of Intake Pressure and Temperature on Auto-Ignition of Fuels with Different Cetane Number and Volatility

DTIC Science & Technology

2012-04-12

1 Effect Of Intake Pressure And Temperature On Auto-Ignition Of Fuels With Different Cetane Number And Volatility C. Jayakumar , U. Joshi, Z...AUTHOR(S) Eric Sattler; C. Jayakumar ; U. Joshi; Z. Zheng; W. Bryzik 5d. PROJECT NUMBER 5e. TASK NUMBER 5f. WORK UNIT NUMBER 7. PERFORMING

X-ray Analysis of Defects and Anomalies in AGR-5/6/7 TRISO Particles

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

Helmreich, Grant W.; Hunn, John D.; Skitt, Darren J.

2017-06-01

Coated particle fuel batches J52O-16-93164, 93165, 93166, 93168, 93169, 93170, and 93172 were produced by Babcock and Wilcox Technologies (BWXT) for possible selection as fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR), or may be used for other tests. Each batch was coated in a 150-mm-diameter production-scale fluidized-bed chemical vapor deposition (CVD) furnace. Tristructural isotropic (TRISO) coatings were deposited on 425-μm-nominal-diameter spherical kernels from BWXT lot J52R-16-69317 containing a mixture of 15.4%-enriched uranium carbide and uranium oxide (UCO), with the exception of Batchmore » 93164, which used similar kernels from BWXT lot J52L-16-69316. The TRISO-coatings consisted of a ~50% dense carbon buffer layer with 100-μmnominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μm-nominal thickness. Each coated particle batch was sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batch was designated by appending the letter A to the end of the batch number (e.g., 93164A). Secondary upgrading by sieving was performed on the upgraded batches to remove specific anomalies identified during analysis for Defective IPyC, and the upgraded batches were designated by appending the letter B to the end of the batch number (e.g., 93165B). Following this secondary upgrading, coated particle composite J52R-16-98005 was produced by BWXT as fuel for the AGR Program’s AGR-5/6/7 irradiation test in the INL ATR. This composite was comprised of coated particle fuel batches J52O-16-93165B, 93168B, 93169B, and 93170B.« less

General classification of ``hot`` particles from the nearest Chernobyl contaminated areas

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

Shabalev, S.I.; Burakov, B.E.; Anderson, E.B.

1997-12-31

The morphology and composition both chemical and radionuclide of the main types of the solid-phase hot particles formed following the accident on the Chernobyl NPP have been studied by SEM, electron microprobe and gamma-spectrometry methods. Differences in many isotopes including: {sup 106}Ru, {sup 134}Cs, {sup 137}Cs dependent upon the hot particle matrix chemical composition was observed. The classification of hot particles based upon the chemical composition of their matrices has been done. It includes three main types: (1) fuel particles with UO{sub x} matrix; (2) fuel-constructional particles with Zr-U-O matrix, (3) hot particles with metallic inclusions of Fe-Cr-Ni. Moreover, theremore » are more rare types of hot particles with silicate or metal matrices. It was shown that only metallic inclusions of Fe-Cr-Ni are concentrators of {sup 106}Ru, which caused this nuclides assimilation in the molten stainless steel during the initial stages of the accident. Soils contamination of non-radioactive lead oxide particles in the Chernobyl NPP region were noticed. It was supposed that part of metallic lead, dropped from helicopters into burning reactor during first days of accident, was evaporated and oxidized accompanying solid oxide particles formation.« less

Processing of uranium oxide and silicon carbide based fuel using polymer infiltration and pyrolysis

NASA Astrophysics Data System (ADS)

Singh, Abhishek K.; Zunjarrao, Suraj C.; Singh, Raman P.

2008-09-01

Ceramic composite pellets consisting of uranium oxide, UO 2, contained within a silicon carbide matrix, were fabricated using a novel processing technique based on polymer infiltration and pyrolysis (PIP). In this process, particles of depleted uranium oxide, in the form of U 3O 8, were dispersed in liquid allylhydridopolycarbosilane (AHPCS), and subjected to pyrolysis up to 900 °C under a continuous flow of ultra high purity argon. The pyrolysis of AHPCS, at these temperatures, produced near-stoichiometric amorphous silicon carbide ( a-SiC). Multiple polymer infiltration and pyrolysis (PIP) cycles were performed to minimize open porosity and densify the silicon carbide matrix. Analytical characterization was conducted to investigate chemical interaction between U 3O 8 and SiC. It was observed that U 3O 8 reacted with AHPCS during the very first pyrolysis cycle, and was converted to UO 2. As a result, final composition of the material consisted of UO 2 particles contained in an a-SiC matrix. The physical and mechanical properties were also quantified. It is shown that this processing scheme promotes uniform distribution of uranium fuel source along with a high ceramic yield of the parent matrix.

Microanalytical X-ray imaging of depleted uranium speciation in environmentally aged munitions residues.

PubMed

Crean, Daniel E; Livens, Francis R; Stennett, Martin C; Grolimund, Daniel; Borca, Camelia N; Hyatt, Neil C

2014-01-01

Use of depleted uranium (DU) munitions has resulted in contamination of the near-surface environment with penetrator residues. Uncertainty in the long-term environmental fate of particles produced by impact of DU penetrators with hard targets is a specific concern. In this study DU particles produced in this way and exposed to the surface terrestrial environment for longer than 30 years at a U.K. firing range were characterized using synchrotron X-ray chemical imaging. Two sites were sampled: a surface soil and a disposal area for DU-contaminated wood, and the U speciation was different between the two areas. Surface soil particles showed little extent of alteration, with U speciated as oxides U3O7 and U3O8. Uranium oxidation state and crystalline phase mapping revealed these oxides occur as separate particles, reflecting heterogeneous formation conditions. Particles recovered from the disposal area were substantially weathered, and U(VI) phosphate phases such as meta-ankoleite (K(UO2)(PO4) · 3H2O) were dominant. Chemical imaging revealed domains of contrasting U oxidation state linked to the presence of both U3O7 and meta-ankoleite, indicating growth of a particle alteration layer. This study demonstrates that substantial alteration of DU residues can occur, which directly influences the health and environmental hazards posed by this contamination.

Modes of Ignition of Powder Layers of Nanocomposite Thermites by Electrostatic Discharge

NASA Astrophysics Data System (ADS)

Monk, I.; Schoenitz, M.; Dreizin, E. L.

2017-01-01

Nanocomposite powders with aluminum as a fuel and oxides of molybdenum, copper, bismuth, and iron as oxidizers were prepared by arrested reactive milling. The powders were placed in 0.5-mm-thick layers and ignited by electrostatic discharge (ESD) in air. In different tests, time-resolved light emission was recorded at 568 nm or in the range of 373-641 nm. The amount of material consumed was recorded as well. Time-resolved temperatures were determined. Two distinct ignition regimes were observed. Prompt ignition occurred within 10 µs of the electric discharge, comparable to what had previously been observed for corresponding powder monolayers. This ignition mode was observed for composites with Bi2O3 and Fe2O3 ignited with a 12-kV discharge, whereas it only occurred at higher spark voltage (20 kV) and energy for CuO and MoO3 composites. Delayed ignition, occurring after 0.1-1 ms following the discharge, was observed for all composites with consistently stronger light emission. Analysis of quenched, partially burned particles showed that the original nanostructure was preserved after prompt ignition but not after delayed ignition. It is proposed that prompt ignition represents direct ESD initiation of composite particles rapidly and adiabatically preheated to high temperatures while keeping the nanostructure intact, resulting in a heterogeneous reaction consuming most of the aluminum. Delayed ignition occurs when particles preheated to lower temperatures start oxidizing at much lower rates, leading to cloud combustion, in which thermal interaction between individual aerosolized burning particles is substantial. During this process, the nanostructure may be lost. Temperature measurements show that nanocomposites with CuO and MoO3 burned superadiabatically with flame temperatures exceeding thermodynamic predictions.

Development and characterization of (Ti, Mo)C carbides reinforced Fe-based surface composite coating produced by laser cladding

NASA Astrophysics Data System (ADS)

Wang, Xinhong; Zhang, Min; Qu, Shiyao

2010-09-01

In this study, in situ multiple carbides reinforced Fe-based surface composite coatings were fabricated successfully by laser cladding a precursor mixture of graphite, ferrotitanium (Fe-Ti) and ferromolybdenum (Fe-Mo) powders. The results showed that (Ti, Mo)C particles with flower-like and cuboidal shapes were in situ formed during the solidification and most shapes of (Ti, Mo)C particles were diversiform according to different contents of Fe-Mo powder in the Fe-Ti-Mo-C system. The growth morphology of the reinforcing (Ti, Mo)C carbide has typically faceted features, indicating that the lateral growth mechanism is still predominant growth mode under rapid solidification conditions. Increasing the amount of Fe-Mo in the reactants led to a decrease of carbide size and an increase of volume fraction of carbides. The coatings had good cracking resistance when the amounts of Fe-Mo were controlled within a range of 15 wt%.

Characterization of Bond Strength of U-Mo Fuel Plates Using the Laser Shockwave Technique: Capabilities and Preliminary Results

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

J. A. Smith; D. L. Cottle; B. H. Rabin

2013-09-01

This report summarizes work conducted to-date on the implementation of new laser-based capabilities for characterization of bond strength in nuclear fuel plates, and presents preliminary results obtained from fresh fuel studies on as-fabricated monolithic fuel consisting of uranium-10 wt.% molybdenum alloys clad in 6061 aluminum by hot isostatic pressing. Characterization involves application of two complementary experimental methods, laser-shock testing and laser-ultrasonic imaging, collectively referred to as the Laser Shockwave Technique (LST), that allows the integrity, physical properties and interfacial bond strength in fuel plates to be evaluated. Example characterization results are provided, including measurement of layer thicknesses, elastic properties ofmore » the constituents, and the location and nature of generated debonds (including kissing bonds). LST provides spatially localized, non-contacting measurements with minimum specimen preparation, and is ideally suited for applications involving radioactive materials, including irradiated materials. The theoretical principles and experimental approaches employed in characterizing nuclear fuel plates are described, and preliminary bond strength measurement results are discussed, with emphasis on demonstrating the capabilities and limitations of these methods. These preliminary results demonstrate the ability to distinguish bond strength variations between different fuel plates. Although additional development work is necessary to validate and qualify the test methods, these results suggest LST is viable as a method to meet fuel qualification requirements to demonstrate acceptable bonding integrity.« less

The MoEDAL Experiment at the LHC - a New Light on the Terascale Frontier

NASA Astrophysics Data System (ADS)

Pinfold, J. L.

2015-07-01

MoEDAL is a pioneering experiment designed to search for highly ionizing avatars of new physics such as magnetic monopoles or massive (pseudo-)stable charged particles. Its groundbreaking physics program defines a number of scenarios that yield potentially revolutionary insights into such foundational questions as: are there extra dimensions or new symmetries; what is the mechanism for the generation of mass; does magnetic charge exist; what is the nature of dark matter; and, how did the big-bang develop. MoEDAL's purpose is to meet such far-reaching challenges at the frontier of the field. The innovative MoEDAL detector employs unconventional methodologies tuned to the prospect of discovery physics. The largely passive MoEDAL detector, deployed at Point 8 on the LHC ring, has a dual nature. First, it acts like a giant camera, comprised of nuclear track detectors - analyzed offline by ultra fast scanning microscopes - sensitive only to new physics. Second, it is uniquely able to trap the particle messengers of physics beyond the Standard Model for further study. MoEDAL's radiation environment is monitored by a state-of-the-art real-time TimePix pixel detector array. A new MoEDAL sub-detector to extend MoEDAL's reach to millicharged, minimally ionizing, particles (MMIPs) is under study.

Low temperature synthesis and sintering of d-UO2 nanoparticles.

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

Nenoff, Tina Maria; Ferreira, Summer Rhodes; Robinson, David B.

We report on the novel room temperature method of synthesizing advanced nuclear fuels; a method that virtually eliminates any volatility of components. This process uses radiolysis to form stable nanoparticle (NP) nuclear transuranic (TRU) fuel surrogates and in-situ heated stage TEM to sinter the NPs. The radiolysis is performed at Sandia's Gamma Irradiation Facility (GIF) 60Co source (3 x 10{sup 6} rad/hr). Using this method, sufficient quantities of fuels for research purposes can be produced for accelerated advanced nuclear fuel development. We are focused on both metallic and oxide alloy nanoparticles of varying compositions, in particular d-U, d-U/La alloys andmore » d-UO2 NPs. We present detailed descriptions of the synthesis procedures, the characterization of the NPs, the sintering of the NPs, and their stability with temperature. We have employed UV-vis, HRTEM, HAADF-STEM imaging, single particle EDX and EFTEM mapping characterization techniques to confirm the composition and alloying of these NPs.« less

Improved Mo-Re VPS Alloys for High-Temperature Uses

NASA Technical Reports Server (NTRS)

Hickman, Robert; Martin, James; McKechnie, Timothy; O'Dell, John Scott

2011-01-01

Dispersion-strengthened molybdenum- rhenium alloys for vacuum plasma spraying (VPS) fabrication of high-temperature-resistant components are undergoing development. In comparison with otherwise equivalent non-dispersion-strengthened Mo-Re alloys, these alloys have improved high-temperature properties. Examples of VPS-fabricated high-temperature-resistant components for which these alloys are expected to be suitable include parts of aircraft and spacecraft engines, furnaces, and nuclear power plants; wear coatings; sputtering targets; x-ray targets; heat pipes in which liquid metals are used as working fluids; and heat exchangers in general. These alloys could also be useful as coating materials in some biomedical applications. The alloys consist of 60 weight percent Mo with 40 weight percent Re made from (1) blends of elemental Mo and Re powders or (2) Re-coated Mo particles that have been subjected to a proprietary powder-alloying-and-spheroidization process. For most of the dispersion- strengthening experiments performed thus far in this development effort, 0.4 volume percent of transition-metal ceramic dispersoids were mixed into the feedstock powders. For one experiment, the proportion of dispersoid was 1 volume percent. In each case, the dispersoid consisted of either ZrN particles having sizes <45 m, ZrO2 particles having sizes of about 1 m, HfO2 particles having sizes <45 m, or HfN particles having sizes <1 m. These materials were chosen for evaluation on the basis of previously published thermodynamic stability data. For comparison, Mo-Re feedstock powders without dispersoids were also prepared.

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.

U.S.-Australia Civilian Nuclear Cooperation: Issues for Congress

DTIC Science & Technology

2010-12-01

Enrichment.......................................................................................................7 Uranium Mining and Milling...Issues for Congress Congressional Research Service 7 The nuclear fuel cycle begins with mining uranium ore and upgrading it to yellowcake. Because...uranium after the mining and milling stage. Commercial enrichment services are available in the United States, Europe, Russia, and Japan. Fuel

Uranium silicide pellet fabrication by powder metallurgy for accident tolerant fuel evaluation and irradiation

DOE PAGES

Harp, Jason Michael; Lessing, Paul Alan; Hoggan, Rita Elaine

2015-06-21

In collaboration with industry, Idaho National Laboratory is investigating uranium silicide for use in future light water reactor fuels as a more accident resistant alternative to uranium oxide base fuels. Specifically this project was focused on producing uranium silicide (U 3Si 2) pellets by conventional powder metallurgy with a density greater than 94% of the theoretical density. This work has produced a process to consistently produce pellets with the desired density through careful optimization of the process. Milling of the U 3Si 2 has been optimized and high phase purity U 3Si 2 has been successfully produced. Results are presentedmore » from sintering studies and microstructural examinations that illustrate the need for a finely ground reproducible particle size distribution in the source powder. The optimized process was used to produce pellets for the Accident Tolerant Fuel-1 irradiation experiment. The average density of these pellets was 11.54 ±0.06 g/cm 3. Additional characterization of the pellets by scaning electron microscopy and X-ray diffraction has also been performed. As a result, pellets produced in this work have been encapsulated for irradiation, and irradiation in the Advanced Test Reactor is expected soon.« less

Uranium silicide pellet fabrication by powder metallurgy for accident tolerant fuel evaluation and irradiation

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

Harp, Jason Michael; Lessing, Paul Alan; Hoggan, Rita Elaine

In collaboration with industry, Idaho National Laboratory is investigating uranium silicide for use in future light water reactor fuels as a more accident resistant alternative to uranium oxide base fuels. Specifically this project was focused on producing uranium silicide (U 3Si 2) pellets by conventional powder metallurgy with a density greater than 94% of the theoretical density. This work has produced a process to consistently produce pellets with the desired density through careful optimization of the process. Milling of the U 3Si 2 has been optimized and high phase purity U 3Si 2 has been successfully produced. Results are presentedmore » from sintering studies and microstructural examinations that illustrate the need for a finely ground reproducible particle size distribution in the source powder. The optimized process was used to produce pellets for the Accident Tolerant Fuel-1 irradiation experiment. The average density of these pellets was 11.54 ±0.06 g/cm 3. Additional characterization of the pellets by scaning electron microscopy and X-ray diffraction has also been performed. As a result, pellets produced in this work have been encapsulated for irradiation, and irradiation in the Advanced Test Reactor is expected soon.« less

Low-pressure hydrocracking of coal-derived Fischer-Tropsch waxes to diesel

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

Dieter Leckel

2007-06-15

Coal-derived low-temperature Fischer-Tropsch (LTFT) wax was hydrocracked at pressures of 3.5-7.0 MPa using silica-alumina-supported sulfided NiW/NiMo and an unsulfided noble metal catalyst, modified with MoO{sub 3}. A low-pressure operation at 3.5 MPa produced a highly isomerized diesel, having low cloud points (from -12 to -28{sup o}C) combined with high cetane numbers (69-73). These properties together with the extremely low sulfur ({lt}5 ppm) and aromatic ({lt}0.5%) contents place coal/liquid (CTL) derived distillates as highly valuable blending components to achieve Eurograde diesel specifications. The upgrading of coal-based LTFT waxes through hydrocracking to high-quality diesel fuel blend components in combination with commercial-feasible coal-integratedmore » gasification combined cycle (coal-IGCC) CO{sub 2} capture and storage schemes should make CTL technology more attractive. 28 refs., 7 figs., 8 tabs.« less

Casting evaluation of U-Zr alloy system fuel slug for SFR prepared by injection casting method

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

Song, Hoon; Kim, Jong-Hwan; Kim, Ki-Hwan

2013-07-01

Metal fuel slugs of U-Pu-Zr alloys for Sodium-cooled Fast Reactor (SFR) have conventionally been fabricated by a vacuum injection casting method. Recently, management of minor actinides (MA) became an important issue because direct disposal of the long-lived MA can be a long-term burden for a tentative repository up to several hundreds of thousand years. In order to recycle transuranic elements (TRU) retained in spent nuclear fuel, remote fabrication capability in a shielded hot cell should be prepared. Moreover, generation of long-lived radioactive wastes and loss of volatile species should be minimized during the recycled fuel fabrication step. In order tomore » prevent the evaporation of volatile elements such as Am, alternative fabrication methods of metal fuel slugs have been studied applying gravity casting, and improved injection casting in KAERI, including melting under inert atmosphere. And then, metal fuel slugs were examined with casting soundness, density, chemical analysis, particle size distribution and microstructural characteristics. Based on these results there is a high level of confidence that Am losses will also be effectively controlled by application of a modest amount of overpressure. A surrogate fuel slug was generally soundly cast by improved injection casting method, melted fuel material under inert atmosphere.« less

Irradiation performance of HTGR recycle fissile fuel

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

Homan, F.J.; Long, E.L. Jr.

1976-08-01

The irradiation performance of candidate HTGR recycle fissile fuel under accelerated testing conditions is reviewed. Failure modes for coated-particle fuels are described, and the performance of candidate recycle fissile fuels is discussed in terms of these failure modes. The bases on which UO/sub 2/ and (Th,U)O/sub 2/ were rejected as candidate recycle fissile fuels are outlined, along with the bases on which the weak-acid resin (WAR)-derived fissile fuel was selected as the reference recycle kernel. Comparisons are made relative to the irradiation behavior of WAR-derived fuels of varying stoichiometry and conclusions are drawn about the optimum stoichiometry and the rangemore » of acceptable values. Plans for future testing in support of specification development, confirmation of the results of accelerated testing by real-time experiments, and improvement in fuel performance and reliability are described.« less

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

NASA Astrophysics Data System (ADS)

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

2016-01-01

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

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

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

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

2016-01-22

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

Developing a laser shockwave model for characterizing diffusion bonded interfaces

NASA Astrophysics Data System (ADS)

Lacy, Jeffrey M.; Smith, James A.; Rabin, Barry H.

2015-03-01

The US National Nuclear Security Agency has a Global Threat Reduction Initiative (GTRI) with the goal of reducing the worldwide use of high-enriched uranium (HEU). A salient component of that initiative is the conversion of research reactors from HEU to low enriched uranium (LEU) fuels. An innovative fuel is being developed to replace HEU in high-power research reactors. The new LEU fuel is a monolithic fuel made from a U-Mo alloy foil encapsulated in Al-6061 cladding. In order to support the fuel qualification process, the Laser Shockwave Technique (LST) is being developed to characterize the clad-clad and fuel-clad interface strengths in fresh and irradiated fuel plates. LST is a non-contact method that uses lasers for the generation and detection of large amplitude acoustic waves to characterize interfaces in nuclear fuel plates. However, because the deposition of laser energy into the containment layer on a specimen's surface is intractably complex, the shock wave energy is inferred from the surface velocity measured on the backside of the fuel plate and the depth of the impression left on the surface by the high pressure plasma pulse created by the shock laser. To help quantify the stresses generated at the interfaces, a finite element method (FEM) model is being utilized. This paper will report on initial efforts to develop and validate the model by comparing numerical and experimental results for back surface velocities and front surface depressions in a single aluminum plate representative of the fuel cladding.

Detailed Analysis of Criteria and Particle Emissions from a Very Large Crude Carrier Using a Novel ECA Fuel.

PubMed

Gysel, Nicholas R; Welch, William A; Johnson, Kent; Miller, Wayne; Cocker, David R

2017-02-07

Ocean going vessels (OGVs) operating within emission control areas (ECA) are required to use fuels with ≤0.1 wt % sulfur. Up to now only distillate fuels could meet the sulfur limits. Recently refiners created a novel low-sulfur heavy-fuel oil (LSHFO) meeting the sulfur limits so questions were posed whether nitric oxide (NO x ) and particulate matter (PM) emissions were the same for the two fuels. This project characterized criteria pollutants and undertook a detailed analysis of PM emissions from a very large crude oil carrier (VLCC) using a distillate ECA fuel (MGO) and novel LSHFO. Results showed emission factors of NO x were ∼5% higher with MGO than LSHFO. PM 2.5 emission factors were ∼3 times higher with LSHFO than MGO, while both were below values reported by Lloyds, U.S. EPA and CARB. A detailed analysis of PM revealed it was >90% organic carbon (OC) for both fuels. Elemental carbon (EC) and soot measured with an AVL microsoot sensor (MSS) reflected black carbon. PM size distributions showed unimodal peaks for both MGO (20-30 nm) and LSHFO (30-50 nm). Particle number (PN) emissions were 28% and 17% higher with the PPS-M compared to the SMPS for LSHFO and MGO, respectively.

Effect of Mo on Microstructures and Wear Properties of In Situ Synthesized Ti(C,N)/Ni-Based Composite Coatings by Laser Cladding.

PubMed

Wu, Fan; Chen, Tao; Wang, Haojun; Liu, Defu

2017-09-06

Using Ni60 alloy, C, TiN and Mo mixed powders as the precursor materials, in situ synthesized Ti(C,N) particles reinforcing Ni-based composite coatings are produced on Ti6Al4V alloys by laser cladding. Phase constituents, microstructures and wear properties of the composite coatings with 0 wt % Mo, 4 wt % Mo and 8 wt % Mo additions are studied comparatively. Results indicate that Ti(C,N) is formed by the in situ metallurgical reaction, the (Ti,Mo)(C,N) rim phase surrounding the Ti(C,N) ceramic particle is synthesized with the addition of Mo, and the increase of Mo content is beneficial to improve the wear properties of the cladding coatings. Because of the effect of Mo, the grains are remarkably refined and a unique core-rim structure that is uniformly dispersed in the matrix appears; meanwhile, the composite coatings with Mo addition exhibit high hardness and excellent wear resistance due to the comprehensive action of dispersion strengthening, fine grain strengthening and solid solution strengthening.

Effect of Mo on Microstructures and Wear Properties of In Situ Synthesized Ti(C,N)/Ni-Based Composite Coatings by Laser Cladding

PubMed Central

Chen, Tao; Wang, Haojun

2017-01-01

Using Ni60 alloy, C, TiN and Mo mixed powders as the precursor materials, in situ synthesized Ti(C,N) particles reinforcing Ni-based composite coatings are produced on Ti6Al4V alloys by laser cladding. Phase constituents, microstructures and wear properties of the composite coatings with 0 wt % Mo, 4 wt % Mo and 8 wt % Mo additions are studied comparatively. Results indicate that Ti(C,N) is formed by the in situ metallurgical reaction, the (Ti,Mo)(C,N) rim phase surrounding the Ti(C,N) ceramic particle is synthesized with the addition of Mo, and the increase of Mo content is beneficial to improve the wear properties of the cladding coatings. Because of the effect of Mo, the grains are remarkably refined and a unique core-rim structure that is uniformly dispersed in the matrix appears; meanwhile, the composite coatings with Mo addition exhibit high hardness and excellent wear resistance due to the comprehensive action of dispersion strengthening, fine grain strengthening and solid solution strengthening. PMID:28878190

In-pile and out-of-pile testing of a molybdenum-uranium dioxide cermet fueled themionic diode

NASA Technical Reports Server (NTRS)

Diianni, D. C.

1972-01-01

The behavior of Mo-UO2 cermet fuel in a diode for thermionic reactor application was studied. The diode had a Mo-0.5 Ti emitter and niobium collector. Output power ranged from 1.4 to 2.8 W/cm squared at emitter and collector temperatures of 1500 deg and 540 C. Thermionic performance was stable within the limits of the instrumentation sensitivity. Through 1000 hours of in-pile operation the emitter was dimensionally stable. However, some fission gases (15 percent) leaked through an inner clad imperfection that occurred during fuel fabrication.

Energy Use and Energy Intensity of U.S. Manufacturing—Data from the 2014 Manufacturing Energy Consumption Survey (MECS)

EIA Publications

2017-01-01

Energy intensity in manufacturing in the United States decreased from 2010 to 2014. U.S. manufacturing overall fuel intensity decreased by 4.4% from 3.016 thousand British thermal units (Btu) per dollar of output in 2010 to 2.882 thousand Btu in 2014.[1] U.S. manufacturing fuel consumption rose 4.7% from 2010 to 2014, although real gross output[2] increased more rapidly at 9.6%.

Redox Conditions and Related Color Change in Eastern Equatorial Pacific Sediments: IODP Site U1334

NASA Astrophysics Data System (ADS)

Kordesch, W. E.; Gussone, N. C.; Hathorne, E. C.; Kimoto, K.; Delaney, M. L.

2011-12-01

This study was prompted by a 65 m thick brown-green color change in deep-sea sediments of IODP Site U1334 (0-38 Ma, 4799 m water depth) that corresponds to its equatorial crossing (caused by the Northward movement of the pacific plate). Green sediment is a visual indicator of reducing conditions in sediment due to enhanced organic matter deposition and burial. Here we use geochemical redox indicators to characterize the effect of equatorial upwelling on bottom water. The modern redox signal is captured in porewater profiles (nitrate, manganese, iron, sulfate) while trace metal Enrichment Factors (EF) in bulk sediment (manganese, uranium, molybdenum, rhenium) normalized to the detrital component (titanium) record redox state at burial. To measure export productivity we also measure biogenic barium. Porewater profiles reveal suboxic diagenesis; profiles follow the expected sequence of nitrate, manganese oxide, and iron oxide reduction with increasing depth. Constant sulfate (~28 μM) implies anoxia has not occurred. Bulk sediment Mn EF are enriched (EF > 1) throughout the record (Mn EF = 15-200) while U and Mo enrichment corresponds to green color and equatorial proximity (U EF = 4-19; Mo EF = 0-7). Constant Mn enrichment implies continuous oxygenation. Uranium and Mo enrichment near the equator represents suboxic conditions also seen in the porewater. Low Re concentrations (below detection) provide additional evidence against anoxia. A comparison of Mn EF from total digestions to samples treated with an additional reductive cleaning step distinguishes between Mn-oxides and Mn-carbonates, indicating oxygenated and reducing conditions respectively. Mn-carbonate occurrence agrees with U and Mo EF; conditions were more reducing near the equator. Bio-Ba shows significant variability over this interval (22-99 mmol g-1). Our geochemical results indicate that bottom waters became suboxic at the equator as a result of equatorial upwelling-influenced increases in organic matter sedimentation. Comparison of results to Site U1335 (0-26 Ma, 4327 m water depth) will test the relative importance of equatorial proximity.

Synthesis of finely divided molybdenum sulfide nanoparticles in propylene carbonate solution

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

Afanasiev, Pavel, E-mail: pavel.afanasiev@ircelyon.univ-lyon1.fr

2014-05-01

Molybdenum sulfide nanoparticles have been prepared from the reflux solution reaction involving ammonium heptamolybdate and elemental sulfur in propylene carbonate. Addition to the reaction mixture of starch as a natural capping agent leads to lesser agglomeration and smaller size of the particles. Nanoparticles of MoS{sub x} (x≈4) of 10–30 nm size are highly divided and form stable colloidal suspensions in organic solvents. Mo K edge EXAFS of the amorphous materials shows rapid exchange of oxygen to sulfur in the molybdenum coordination sphere during the solution reaction. Thermal treatment of the amorphous sulfides MoS{sub x} under nitrogen or hydrogen flow atmore » 400 °C allows obtaining mesoporous MoS{sub 2} materials with very high pore volume and specific surface area, up to 0.45 cm{sup 3}/g and 190 m{sup 2}/g, respectively. The new materials show good potential for the application as unsupported hydrotreating catalysts. - Graphical abstract: Solution reaction in propylene carbonate allows preparing weakly agglomerated molybdenum sulfide with particle size 20 nm and advantageous catalytic properties. - Highlights: • Solution reaction in propylene carbonate yields MoS{sub x} particles near 20 nm size. • Addition of starch as capping agent reduces particles size and hinder agglomeration. • EXAFS at Mo K edge shows rapid oxygen to sulfur exchange in the solution. • Thermal treatment leads to MoS{sub 2} with very high porosity and surface area.« less

Effect of 3-D magnetic fields on neutral particle fueling and exhaust in MAST

NASA Astrophysics Data System (ADS)

Flesch, Kurt; Kremeyer, Thierry; Waters, Ian; Schmitz, Oliver; Kirk, Andrew; Harrison, James

2017-10-01

The application of resonant magnetic perturbations (RMPs) is used to suppress edge localized modes but causes in many cases a density pump-out. At MAST, this particle pump out was found to be connected to an amplifying MHD plasma response. An analysis is presented on past MAST discharges to understand the effect of these RMPs on the neutral household and on changes in neutral fueling and exhaust during the pump out. A global, 0-D particle balance model was used to study the neutral dynamics and plasma confinement during shots with and without RMP application. Using the D α emission measured by filterscopes and a calibrated 1-D CCD camera, as well as S/XB coefficients determined by the edge plasma parameters, globally averaged ion confinement times were calculated. In L-mode, discharges with RMPs that caused an MHD response had a 15-20% decrease in confinement time but an increase in total recycling flux. The application of RMPs in H-mode caused either a decrease or no change in confinement, like those in L-mode, depending on the configuration of the RMPs and plasma response. A spectroscopically assisted Penning gauge is being prepared for the next campaign at MAST-U to extend this particle balance to study impurity exhaust with RMPs. This work was funded in part by the U.S. DoE under Grant DE-SC0012315.

Experimental study of the rearrangements of valence protons and neutrons amongst single-particle orbits during double-β decay in 100Mo

NASA Astrophysics Data System (ADS)

Freeman, S. J.; Sharp, D. K.; McAllister, S. A.; Kay, B. P.; Deibel, C. M.; Faestermann, T.; Hertenberger, R.; Mitchell, A. J.; Schiffer, J. P.; Szwec, S. V.; Thomas, J. S.; Wirth, H.-F.

2017-11-01

The rearrangements of protons and neutrons amongst the valence single-particle orbitals during double-β decay of 100Mo have been determined by measuring cross sections in (d ,p ), (p ,d ), (3He,α ), and (3He,d ) reactions on Mo,10098 and Ru,102100 targets. The deduced nucleon occupancies reveal significant discrepancies when compared with theoretical calculations; the same calculations have previously been used to determine the nuclear matrix element associated with the decay probability of double-β decay of the 100Mo system.

Exploratory analysis of the potential relationship between urinary molybdenum and bone mineral density among adult men and women from NHANES 2007-2010.

PubMed

Lewis, Ryan C; Johns, Lauren E; Meeker, John D

2016-12-01

Human exposure to molybdenum (Mo) may play a role in reducing bone mineral density (BMD) by interfering with steroid sex hormone levels. To begin to address gaps in the literature on this topic, the potential relationship between urinary Mo (U-Mo) and BMD at the femoral neck (FN-BMD) and lumbar spine (LS-BMD) was explored in a sample of 1496 adults participating in the 2007-2010 cycles of the National Health and Nutrition Examination Survey. Associations were assessed using multiple linear regression models stratified on sex and age. In adjusted models for 50-80+ year-old women, there was a statistically significant inverse relationship between natural log-U-Mo and LS-BMD (p-value: 0.002), and a statistically significant dose-dependent decrease in LS-BMD with increasing U-Mo quartiles (trend p-value: 0.002). A suggestive (trend p-value: 0.08), dose-dependent decrease in FN-BMD with increasing U-Mo quartiles was noted in this group of women as well. All other adjusted models revealed no statistically significant or suggestive relationships between U-Mo and FN-BMD or LS-BMD. Bone health is important for overall human health and well-being and, given the exploratory nature of this work, additional studies are needed to confirm the results in other populations, and clarify the potential underlying mechanisms of Mo on BMD. Copyright © 2016 Elsevier Ltd. All rights reserved.

Comparison of fresh fuel experimental measurements to MCNPX calculations using self-interrogation neutron resonance densitometry

NASA Astrophysics Data System (ADS)

LaFleur, Adrienne M.; Charlton, William S.; Menlove, Howard O.; Swinhoe, Martyn T.

2012-07-01

A new non-destructive assay technique called Self-Interrogation Neutron Resonance Densitometry (SINRD) is currently being developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for Light Water Reactor (LWR) fuel assemblies. SINRD consists of four 235U fission chambers (FCs): bare FC, boron carbide shielded FC, Gd covered FC, and Cd covered FC. Ratios of different FCs are used to determine the amount of resonance absorption from 235U in the fuel assembly. The sensitivity of this technique is based on using the same fissile materials in the FCs as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n,f) reaction peaks in the fission chamber. In this work, experimental measurements were performed in air with SINRD using a reference Pressurized Water Reactor (PWR) 15×15 low enriched uranium (LEU) fresh fuel assembly at LANL. The purpose of this experiment was to assess the following capabilities of SINRD: (1) ability to measure the effective 235U enrichment of the PWR fresh LEU fuel assembly and (2) sensitivity and penetrability to the removal of fuel pins from an assembly. These measurements were compared to Monte Carlo N-Particle eXtended transport code (MCNPX) simulations to verify the accuracy of the MCNPX model of SINRD. The reproducibility of experimental measurements via MCNPX simulations is essential to validating the results and conclusions obtained from the simulations of SINRD for LWR spent fuel assemblies.

Thorium-232 fission induced by light charged particles up to 70 MeV

NASA Astrophysics Data System (ADS)

Métivier, Vincent; Duchemin, Charlotte; Guertin, Arnaud; Michel, Nathalie; Haddad, Férid

2017-09-01

Studies have been devoted to the production of alpha emitters for medical application in collaboration with the GIP ARRONAX that possesses a high energy and high intensity multi-particle cyclotron. The productions of Ra-223, Ac-225 and U-230 have been investigated from the Th-232(p,x) and Th-232(d,x) reactions using the stacked-foils method and gamma spectrometry measurements. These reactions have led to the production of several fission products, including some with a medical interest like Mo-99, Cd-115g and I-131. This article presents cross section data of fission products obtained from these undedicated experiments. These data have been also compared with the TALYS code results.

Photocatalytic degradation of RhB with microwave prepared PbMoO4.

PubMed

Hernández-Uresti, Diana B; Aguilar-Garib, Juan A; Martínez-de la Cruz, Azael

2012-01-01

Synthesized PbMoO4 from H2MoO4 and Pb(NO3)2 with microwaves was compared, in terms of its photocatalytic activity as catalyzer for decomposing rhodamine B (RhB), against samples prepared by hydrothermal and sonochemical methods from the same precursors. Microwave synthesis lasted 20 minutes; hydrothermal, 10 minutes and sonochemical method, 1 hour. Xrays diffraction patterns show that PbMoO4 prepared by these three routes is compounded by the same phase. It is found that microwave synthesized PbMoO4 particles are rounder, in an intermediate size (250 nm), compared to sonochemical (100 nm) and hydrothermal (500 nm) routes; microwave particles also exhibit higher photocatalytic activity for degradation of RhB under a xenon lamp. This difference is not explicable in terms of surface area measurements, but could be explained by UV Light scattering by the rounder particles produced by means of the microwave processing, which are about one half size compared to the wavelength.

Transformation of Cerium Oxide Nanoparticles from a Diesel Fuel Additive during Combustion in a Diesel Engine.

PubMed

Dale, James G; Cox, Steven S; Vance, Marina E; Marr, Linsey C; Hochella, Michael F

2017-02-21

Nanoscale cerium oxide is used as a diesel fuel additive to reduce particulate matter emissions and increase fuel economy, but its fate in the environment has not been established. Cerium oxide released as a result of the combustion of diesel fuel containing the additive Envirox, which utilizes suspended nanoscale cerium oxide to reduce particulate matter emissions and increase fuel economy, was captured from the exhaust stream of a diesel engine and was characterized using a combination of bulk analytical techniques and high resolution transmission electron microscopy. The combustion process induced significant changes in the size and morphology of the particles; ∼15 nm aggregates consisting of 5-7 nm faceted crystals in the fuel additive became 50-300 nm, near-spherical, single crystals in the exhaust. Electron diffraction identified the original cerium oxide particles as cerium(IV) oxide (CeO 2 , standard FCC structure) with no detectable quantities of Ce(III), whereas in the exhaust the ceria particles had additional electron diffraction reflections indicative of a CeO 2 superstructure containing ordered oxygen vacancies. The surfactant coating present on the cerium oxide particles in the additive was lost during combustion, but in roughly 30% of the observed particles in the exhaust, a new surface coating formed, approximately 2-5 nm thick. The results of this study suggest that pristine, laboratory-produced, nanoscale cerium oxide is not a good substitute for the cerium oxide released from fuel-borne catalyst applications and that future toxicity experiments and modeling will require the use/consideration of more realistic materials.

Self Regulating Fiber Fuel Cell

DTIC Science & Technology

2010-08-16

12000 68.2 77.4 24/7 Extreme Rigid liquid hydrogen fuel cell Medis 68 X 97 X 57 20000 53.2 108.1 Fiber Fuel Cell Flexible Individual fiber Honeywell...which allows hydrogen and water vapor to permeate freely but prevents liquids from entering or fuel particles from escaping. The SPM permeability...S is the solubility and D is the diffusivity. Solubility and diffusivity data vs. pressure for hydrogen in Nafion is not available in the literature

Sales of Fossil Fuels Produced from Federal and Indian Lands, FY 2003 through FY 2014

EIA Publications

2015-01-01

The U.S. Energy Information Administration estimates that total sales of fossil fuels produced from Federal and Indian Lands increased in fiscal year 2014 compared to fiscal year 2013. Production of crude oil increased 7%, natural gas production declined 7%, natural gas plant liquids production increased by 8%, and coal production increased slightly. Detailed tables and maps of production, by State, are contained in the report. EIA’s estimates are based on data provided by the U.S. Department of the Interior’s Office of Natural Resources Revenue.

Numerical characterization of micro-cell UO2sbnd Mo pellet for enhanced thermal performance

NASA Astrophysics Data System (ADS)

Lee, Heung Soo; Kim, Dong-Joo; Kim, Sun Woo; Yang, Jae Ho; Koo, Yang-Hyun; Kim, Dong Rip

2016-08-01

Metallic micro-cell UO2 pellet with high thermal conductivity has received attention as a promising accident-tolerant fuel. Although experimental demonstrations have been successful, studies on the potency of current metallic micro-cell UO2 fuels for further enhancement of thermal performance are lacking. Here, we numerically investigated the thermal conductivities of micro-cell UO2sbnd Mo pellets in terms of the amount of Mo content, the unit cell size, and the aspect ratio of the micro-cells. The results showed good agreement with experimental measurements, and more importantly, indicated the importance of optimizing the unit cell geometries of the micro-cell pellets for greater increases in thermal conductivity. Consequently, the micro-cell UO2sbnd Mo pellets (5 vol% Mo) with modified geometries increased the thermal conductivity of the current UO2 pellets by about 2.5 times, and lowered the temperature gradient within the pellets by 62.9% under a linear heat generation rate of 200 W/cm.

Isotopic signature and nano-texture of cesium-rich micro-particles: Release of uranium and fission products from the Fukushima Daiichi Nuclear Power Plant.

PubMed

Imoto, Junpei; Ochiai, Asumi; Furuki, Genki; Suetake, Mizuki; Ikehara, Ryohei; Horie, Kenji; Takehara, Mami; Yamasaki, Shinya; Nanba, Kenji; Ohnuki, Toshihiko; Law, Gareth T W; Grambow, Bernd; Ewing, Rodney C; Utsunomiya, Satoshi

2017-07-14

Highly radioactive cesium-rich microparticles (CsMPs) released from the Fukushima Daiichi Nuclear Power Plant (FDNPP) provide nano-scale chemical fingerprints of the 2011 tragedy. U, Cs, Ba, Rb, K, and Ca isotopic ratios were determined on three CsMPs (3.79-780 Bq) collected within ~10 km from the FDNPP to determine the CsMPs' origin and mechanism of formation. Apart from crystalline Fe-pollucite, CsFeSi 2 O 6  · nH 2 O, CsMPs are comprised mainly of Zn-Fe-oxide nanoparticles in a SiO 2 glass matrix (up to ~30 wt% of Cs and ~1 wt% of U mainly associated with Zn-Fe-oxide). The 235 U/ 238 U values in two CsMPs: 0.030 (±0.005) and 0.029 (±0.003), are consistent with that of enriched nuclear fuel. The values are higher than the average burnup estimated by the ORIGEN code and lower than non-irradiated fuel, suggesting non-uniform volatilization of U from melted fuels with different levels of burnup, followed by sorption onto Zn-Fe-oxides. The nano-scale texture and isotopic analyses provide a partial record of the chemical reactions that occurred in the fuel during meltdown. Also, the CsMPs were an important medium of transport for the released radionuclides in a respirable form.

MoEDAL - a new light on the high-energy frontier

NASA Astrophysics Data System (ADS)

Fairbairn, Malcolm; Pinfold, James L.

2017-01-01

In 2010, the MoEDAL (MOnopole and Exotics Detector at the LHC) experiment at the Large Hadron Collider (LHC) was unanimously approved by European Centre for Nuclear Research's Research Board to start data taking in 2015. MoEDAL is a pioneering experiment designed to search for highly ionising manifestations of new physics such as magnetic monopoles or massive (pseudo-)stable charged particles. Its groundbreaking physics programme defines a number of scenarios that yield potentially revolutionary insights into such foundational questions as: are there extra dimensions or new symmetries; does magnetic charge exist; what is the nature of dark matter; and, how did the Big Bang develop. MoEDAL's purpose is to meet such far-reaching challenges at the frontier of the field. The innovative MoEDAL detector employs unconventional methodologies tuned to the prospect of discovery physics. The largely passive MoEDAL detector, deployed at Point 8 on the LHC ring, has a dual nature. First, it acts like a giant camera, comprised of nuclear track detectors - analysed offline by ultra fast scanning microscopes - sensitive only to new physics. Second, it is uniquely able to trap the particle messengers of physics beyond the Standard Model for further study. MoEDAL's radiation environment is monitored by a state-of-the-art real-time TimePix pixel detector array. A new MoEDAL sub-detector designed to extend MoEDAL reach to mini-charged, minimally ionising particles is under study.

The degree of peri-implant osteolysis induced by PEEK, CoCrMo, and HXLPE wear particles: a study based on a porous Ti6Al4V implant in a rabbit model.

PubMed

Du, Zhe; Zhu, Zhonglin; Wang, You

2018-01-31

Polyether-ether-ketone (PEEK), cobalt-chromium-molybdenum (CoCrMo), and highly cross-linked polyethylene (HXLPE) are biomaterials used in orthopedic implants; their wear particles are considered to induce peri-implant osteolysis. We examined whether different particle types induce the same degree of peri-implant osteolysis. Forty female rabbits were randomly divided into four groups-the control group (n = 10), which received implantation operation and sham operation at 1 month postoperation; three experimental groups (n = 10 in each group), which received implantation operation along with administration of 0.1 mL of particle suspension (approximately 1.0 × 10 8 PEEK, CoCrMo, or HXLPE wear particles) into the knee joint at 1 month postoperation. All rabbits were sacrificed at 2 months postoperation. The synovium was removed and histologically assessed. The distal femurs with the implants were analyzed via micro-computed tomography (CT) and hard tissue biopsy. The average size of almost 90% of the particles was < 5 μm, indicating no significant difference in the three particle types. IL-1β, IL-8, TNFα, RANKL, and MCP-1 expression in PEEK and CoCrMo groups was high, while that in the HXLPE group was low. The bone density (BD) and bone volume/total volume (BV/TV) of the porous structures (part of the implants in all groups) in experimental groups did not decrease markedly (p > 0.05), while BD in the peripheral regions in experimental groups decreased markedly compared to control groups (p < 0.05). BV/TV in the peripheral regions was significantly decreased in PEEK and CoCrMo groups when compared to control group (p < 0.05), while no significant difference was noted between HXLPE and control groups (p > 0.05). The changes in BV observed in the hard tissue sections were consistent with those noted in the micro-CT findings. PEEK, CoCrMo, and HXLPE wear particles (approximately having the same size and doses) induce peri-implant osteolysis to a different degree: HXLPE particles induce peri-implant osteolysis to a mild degree, while PEEK and CoCrMo particles caused significant peri-implant osteolysis. In case of a porous implant, osteolysis occurred primarily in the peripheral region, rather than in the porous structures. Our findings would be helpful for implant designers to choose friction pairs in orthopedic components.

Normal and grazing incidence pulsed laser deposition of nanostructured MoSx hydrogen evolution catalysts from a MoS2 target

NASA Astrophysics Data System (ADS)

Fominski, V. Yu.; Romanov, R. I.; Fominski, D. V.; Dzhumaev, P. S.; Troyan, I. A.

2018-06-01

Pulsed laser ablation of a MoS2 target causes enhanced splashing of the material. So, for MoSx films obtained by pulsed laser deposition (PLD) in the conventional normal incidence (NI) configuration, their typical morphology is characterized by an underlying granular structure with an overlayer of widely dispersed spherical Mo and MoSx particles possessing micro-, sub-micro- and nanometer sizes. We investigated the possibility of using high surface roughness, which occurs due to particle deposition, as a support with a large exposed surface area for thin MoSx catalytic layers for the hydrogen evolution reaction (HER). For comparison, the HER performance of MoSx layers formed by grazing incidence (GI) PLD was studied. During GI-PLD, a substrate was placed along the direction of laser plume transport and few large particles loaded the substrate. The local structure and composition of thin MoSx layers formed by the deposition of the vapor component of the laser plume were varied by changing the pressure of the buffer gas (argon, Ar). In the case of NI-PLD, an increase in Ar pressure caused the formation of quasi-amorphous MoSx (x ≥ 2) films that possessed highly active catalytic sites on the edges of the layered MoS2 nanophase. At the same time, a decrease in the deposition rate of the MoSx film appeared due to the scattering of the vapor flux by Ar molecules during flux transport from the target to the substrate. This effect prevented uniform deposition of the MoSx catalytic film on the surface of most particles, whose deposition rate was independent of Ar pressure. The scattered vapor flux containing Mo and S atoms was a dominant source for MoSx film growth during GI-PLD. The thickness and composition distribution of the MoSx film on the substrate depended on both the pressure of the buffer gas and the distance from the target. For 1.0-2.5 cm from the target, the deposition rate was quite sufficient to form S-enriched quasi-amorphous MoSx (2.5 < x < 6) catalytic films that consisted of densely packed 30-50 nm nanoparticles. The GI-PLD films possessed a greater density of catalytically active sites with a distinct local atomic configuration including edge sites of the layered MoS2 nanophase and diverse S ligands in the amorphous phase, which contained Mo3-S clusters. At a modest loading of ∼300 μg/cm2 on glassy carbon substrates and an overpotential of -140 mV, these films activated H2 production with geometric current densities up to -10 mA/cm2.

Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

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

Renfro, David G; Chandler, David; Cook, David Howard

2014-11-01

Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully convertedmore » using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the complex aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The present studies used current analytical tools to evaluate the various alternate designs for cycle length, scientific performance (e.g., neutron scattering), and steady-state and transient thermal performance using both safety limit and nominal parameter assumptions. The studies concluded that a new reference design combining a permanent absorber in the lower unfueled region of all of the fuel plates, a burnable absorber in the inner element side plates, and a relocated and reshaped (but still radially contoured) fuel zone will allow successful conversion of HFIR. Future collaboration with the program will reveal whether the new reference design can be fabricated reliably and affordably. Following this feedback, additional studies using state-of-the-art developmental analytical tools are proposed to optimize the design of the fuel zone radial contour and the amount and location of both types of neutron absorbers to further flatten thermal peaks while maximizing the performance of the reactor.« less

The MoEDAL experiment at the LHC. Searching beyond the standard model

NASA Astrophysics Data System (ADS)

Pinfold, James L.

2016-11-01

MoEDAL is a pioneering experiment designed to search for highly ionizing avatars of new physics such as magnetic monopoles or massive (pseudo-)stable charged particles. Its groundbreaking physics program defines a number of scenarios that yield potentially revolutionary insights into such foundational questions as: are there extra dimensions or new symmetries; what is the mechanism for the generation of mass; does magnetic charge exist; what is the nature of dark matter; and, how did the big-bang develop. MoEDAL's purpose is to meet such far-reaching challenges at the frontier of the field. The innovative MoEDAL detector employs unconventional methodologies tuned to the prospect of discovery physics. The largely passive MoEDAL detector, deployed at Point 8 on the LHC ring, has a dual nature. First, it acts like a giant camera, comprised of nuclear track detectors - analyzed offline by ultra fast scanning microscopes - sensitive only to new physics. Second, it is uniquely able to trap the particle messengers of physics beyond the Standard Model for further study. MoEDAL's radiation environment is monitored by a state-of-the-art real-time TimePix pixel detector array. A new MoEDAL sub-detector to extend MoEDAL's reach to millicharged, minimally ionizing, particles (MMIPs) is under study Finally we shall describe the next step for MoEDAL called Cosmic MoEDAL, where we define a very large high altitude array to take the search for highly ionizing avatars of new physics to higher masses that are available from the cosmos.

75 FR 70947 - Notice of Lodging of Proposed Consent Decree Under the Resource Conservation and Recovery Act

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-19

..., Kansas City, MO 64106, and at the Environmental Protection Agency, Region 7, 901 N. 5th St., Kansas City..., or 37.25 (if attachments are requested) (25 cents per page reproduction cost) payable to the U.S...

Microstructure and mechanical behavior of Zr substrates coated with FeCrAl and Mo by cold-spraying

NASA Astrophysics Data System (ADS)

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

2018-06-01

FeCrAl and Mo layers were cold-sprayed onto a Zr surface, with the Mo layer introduced between the FeCrAl coating and the Zr matrix preventing high-temperature interdiffusion. Microstructural characterization of the first-deposited Mo layer and the Zr matrix immediately below the Mo/Zr interface was performed using transmission electron microscopy, and near-interface elemental distributions were obtained using energy-dispersive X-ray spectroscopy. The deformation of the coated Mo powder induced the formation of microbands and mechanically interlocked nanoscale structures. The mechanical behavior of Zr with a coating layer was compared with those characteristic of conventional Zr samples. The coated sample showed smaller strength reduction in the test conducted at elevated temperature. The hardness and fracture morphology of the Zr matrix near the interface region were investigated to determine the effect of impacting Mo particles on the matrix microstructure. The enhanced hardness and cleavage fracture morphology of the Zr matrix immediately below the Mo/Zr interface indicated the occurrence of localized deformation owing to Mo particle impact.

Process R&D for Particle Size Control of Molybdenum Oxide

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

Sen, Sujat; Dzwiniel, Trevor; Pupek, Krzysztof

The primary goal of this study was to produce MoO 3 powder with a particle size range of 50 to 200 μm for use in targets for production of the medical isotope 99Mo. Molybdenum metal powder is commercially produced by thermal reduction of oxides in a hydrogen atmosphere. The most common source material is MoO 3, which is derived by the thermal decomposition of ammonium heptamolybdate (AHM). However, the particle size of the currently produced MoO 3 is too small, resulting in Mo powder that is too fine to properly sinter and press into the desired target. In this study,more » effects of heating rate, heating temperature, gas type, gas flow rate, and isothermal heating were investigated for the decomposition of AHM. The main conclusions were as follows: lower heating rate (2-10°C/min) minimizes breakdown of aggregates, recrystallized samples with millimeter-sized aggregates are resistant to various heat treatments, extended isothermal heating at >600°C leads to significant sintering, and inert gas and high gas flow rate (up to 2000 ml/min) did not significantly affect particle size distribution or composition. In addition, attempts to recover AHM from an aqueous solution by several methods (spray drying, precipitation, and low temperature crystallization) failed to achieve the desired particle size range of 50 to 200 μm. Further studies are planned.« less

Preparation and evaluation of advanced electrocatalysts for phosphoric acid fuel cells

NASA Technical Reports Server (NTRS)

Stonehart, P.; Baris, J.; Hochmuth, J.; Pagliaro, P.

1981-01-01

A number of electrocatalyst combinations were prepared and characterized. These electrocatalysts were formulated to contain platinum combined with transition metal carbide forming elements (W, Mo, V) for cathodes and platinum combined with palladium for anodes. High resolution electron microscopy was used to determine the crystallite size and dispersion of platinum-palladium alloy electrocatalysts in order to provide analytical support for the electrochemical determinations of the particle dispersions. An equation was derived which correlates palladium crystallite size with electrochemical hydrogen adsorption. Based on comparisons of electrocatalyst performances in the presence of pure hydrogen and hydrogen containing carbon monoxide, it was shown that the apparent poisoning of the electrocatalyst by carbon monoxide is influenced by the electrode structure.

Scalability of the LEU-Modified Cintichem Process: 3-MeV Van de Graaff and 35-MeV Electron Linear Accelerator Studies

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

Rotsch, David A.; Brossard, Tom; Roussin, Ethan

Molybdenum-99, the mother of Tc-99m, can be produced from fission of U-235 in nuclear reactors and purified from fission products by the Cintichem process, later modified for low-enriched uranium (LEU) targets. The key step in this process is the precipitation of Mo with α-benzoin oxime (ABO). The stability of this complex to radiation has been examined. Molybdenum-ABO was irradiated with 3 MeV electrons produced by a Van de Graaff generator and 35 MeV electrons produced by a 50 MeV/25 kW electron linear accelerator. Dose equivalents of 1.7–31.2 kCi of Mo-99 were administered to freshly prepared Mo-ABO. Irradiated samples of Mo-ABOmore » were processed according to the LEU Modified-Cintichem process. The Van de Graaff data indicated good radiation stability of the Mo-ABO complex up to ~15 kCi dose equivalents of Mo-99 and nearly complete destruction at doses >24 kCi Mo-99. The linear accelerator data indicate that even at 6.2 kCi of Mo-99 equivalence of dose, the sample lost ~20% of Mo-99. The 20% loss of Mo-99 at this low dose may be attributed to thermal decomposition of the product from the heat deposited in the sample during irradiation.« less

Transportation fuel production by combination of LDPE thermal cracking and catalytic hydroreforming

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

Escola, J.M., E-mail: josemaria.escola.saez@urjc.es; Aguado, J.; Serrano, D.P.

2014-11-15

Highlights: • h-Beta samples were impregnated with Ni nitrate to achieve Ni contents of 1.5%, 4%, 7% and 10%. • Larger and more easily reducible metal particles were obtained on Ni 7%/h-Beta and Ni 10%/h-Beta. • Higher Ni contents increased the amount of gases at the expenses of diesel fractions. • Maximum selectivity to automotive fuels (∼81%) was obtained with Ni 7%/h-Beta. • Ni loading also enhanced olefins saturation up to Ni 7%/h-Beta. - Abstract: Fuel production from plastics is a promising way to reduce landfilling rates while obtaining valuable products. The usage of Ni-supported hierarchical Beta zeolite (h-Beta) formore » the hydroreforming of the oils coming from LDPE thermal cracking has proved to produce high selectivities to gasoline and diesel fuels (>80%). In the present work, the effect of the Ni loading on Ni/h-Beta is investigated in the hydroreforming of the oils form LDPE thermal cracking. h-Beta samples were impregnated with Ni nitrate, calcined and reduced in H{sub 2} up to 550 °C to achieve different Ni contents: 1.5%, 4%, 7% and 10%. Larger and more easily reducible metal particles were obtained on Ni 7%/h-Beta and Ni 10%/h-Beta. Hydroreforming tests were carried out in autoclave reactor at 310 °C, under 20 bar H{sub 2}, for 45 min. Ni content progressively increased the amount of gases at the expenses of diesel fractions, while gasoline remained approximately constant about 52–54%. Maximum selectivity to automotive fuels (∼81%) was obtained with Ni 7%/h-Beta. Ni loading also enhanced olefins saturation up to Ni 7%/h-Beta. High cetane indices (71–86) and octane numbers (89–91) were obtained over all the catalysts. Regarding the different studied Ni contents, Ni 7%/h-Beta constitutes a rather promising catalyst for obtaining high quality fuels from LDPE thermal cracking oils.« less

A summary of the results from the DOE advanced gas reactor (AGR) fuel development and qualification program

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

Petti, David Andrew

2017-04-01

Modular high temperature gas-cooled reactor (HTGR) designs were developed to provide natural safety, which prevents core damage under all licensing basis events. The principle that guides their design concepts is to passively maintain core temperatures below fission product release thresholds under all accident scenarios. The required level of fuel performance and fission product retention reduces the radioactive source term by many orders of magnitude relative to source terms for other reactor types and allows a graded approach to emergency planning and the potential elimination of the need for evacuation and sheltering beyond a small exclusion area. Achieving this level, however,more » is predicated on exceptionally high coated-particle fuel fabrication quality and excellent performance under normal operation and accident conditions. The design goal of modular HTGRs is to meet the Environmental Protection Agency (EPA) Protective Action Guides (PAGs) for offsite dose at the Exclusion Area Boundary (EAB). To achieve this, the reactor design concepts require a level of fuel integrity that is far better than that achieved for all prior U.S.-manufactured tristructural isotropic (TRISO) coated particle fuel.« less

Prognosis and comparison of performances of composite CERCER and CERMET fuels dedicated to transmutation of TRU in an EFIT ADS

NASA Astrophysics Data System (ADS)

Sobolev, V.; Uyttenhove, W.; Thetford, R.; Maschek, W.

2011-07-01

The neutronic and thermomechanical performances of two composite fuel systems: CERCER with (Pu,Np,Am,Cm)O 2-x fuel particles in ceramic MgO matrix and CERMET with metallic Mo matrix, selected for transmutation of minor actinides in the European Facility for Industrial Transmutation (EFIT), were analysed aiming at their optimisation. The ALEPH burnup code system, based on MNCPX and ORIGEN codes and JEFF3.1 nuclear data library, and the modern version of the fuel rod performance code TRAFIC were used for this analysis. Because experimental data on the properties of the mixed minor-actinide oxides are scarce, and the in-reactor behaviour of the T91 steel chosen as cladding, as well as of the corrosion protective layer, is still not well-known, a set of "best estimates" provided the properties used in the code. The obtained results indicate that both fuel candidates, CERCER and CERMET, can satisfy the fuel design and safety criteria of EFIT. The residence time for both types of fuel elements can reach about 5 years with the reactivity swing within ±1000 pcm, and about 22% of the loaded MA is transmuted during this period. However, the fuel centreline temperature in the hottest CERCER fuel rod is close to the temperature above which MgO matrix becomes chemically instable. Moreover, a weak PCMI can appear in about 3 years of operation. The CERMET fuel can provide larger safety margins: the fuel temperature is more than 1000 K below the permitted level of 2380 K and the pellet-cladding gap remains open until the end of operation.

RECONNAISSANCE FOR URANIUM IN ASPHALT-BEARING ROCKS IN THE WESTERN UNITED STATES

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

Hail, W.J. Jr.

1957-01-01

An appraisal of asphait-bearing rocks as potential sources of uranium was made during 1953 and 1954 in 45 areas in Calif., Utah, Wyo., Mont., N. Mex., Tex., Okla., and Mo. A total of 202 samples from these areas was analyzed for uranium. The oldest rocks sampled are Ordovician in age, and the youngest are Recent. Although none of the deposits are of value at this time as a source of U, some of the deposits may constitute a low-grade U resource, but recovery of the U will depend upon the primary use of the asphalt. Significant amounts of U lnmore » the ash of oil extracted from these rocks were found in samples from 7 of the 45 areas examined. These areas are Chalome Creek, McKittrick, Edna, and Los Alamos Calif.; Vernal, Utah; Sulphur, Okla.; and Ellis, Mo. The average U content in the ash of the extracted oil of samples from these 7 areas ranges from 0.028 to 0.376%. All except the Chalone Creek area contain large estimated reserves of asphalt-bearing rock, ranging from 15 million to almost 2 billion tons. The average U content of samples from 13 additiomal areas ranges from 0.020 to 0.06B% in the ash of the extracted oil. Many of these areas contain very large reserves of asphalt-bearing rocks. It is believed that most of the asphalt deposits are oil residues, and that the U was introduced during or after the late stages of oil movement and loss of the lighter oil fractions. (auth)« less

Characterization of wear debris from metal-on-metal hip implants during normal wear versus edge-loading conditions.

PubMed

Kovochich, Michael; Fung, Ernest S; Donovan, Ellen; Unice, Kenneth M; Paustenbach, Dennis J; Finley, Brent L

2018-04-01

Advantages of second-generation metal-on-metal (MoM) hip implants include low volumetric wear rates and the release of nanosized wear particles that are chemically inert and readily cleared from local tissue. In some patients, edge loading conditions occur, which result in higher volumetric wear. The objective of this study was to characterize the size, morphology, and chemistry of wear particles released from MoM hip implants during normal (40° angle) and edge-loading (65° angle with microseparation) conditions. The mean primary particle size by volume under normal wear was 35 nm (range: 9-152 nm) compared with 95 nm (range: 6-573 nm) under edge-loading conditions. Hydrodynamic diameter analysis by volume showed that particles from normal wear were in the nano- (<100 nm) to submicron (<1000 nm) size range, whereas edge-loading conditions generated particles that ranged from <100 nm up to 3000-6000 nm in size. Particles isolated from normal wear were primarily chromium (98.5%) and round to oval in shape. Edge-loading conditions generated more elongated particles (4.5%) (aspect ratio ≥ 2.5) and more CoCr alloy particles (9.3%) compared with normal wear conditions (1.3% CoCr particles). By total mass, edge-loading particles contained approximately 640-fold more cobalt than normal wear particles. Our findings suggest that high wear conditions are a potential risk factor for adverse local tissue effects in MoM patients who experience edge loading. This study is the first to characterize both the physical and chemical characteristics of MoM wear particles collected under normal and edge-loading conditions. © 2017 Wiley Periodicals, Inc. J Biomed Mater Res Part B: Appl Biomater, 106B: 986-996, 2018. © 2017 Wiley Periodicals, Inc.

Novel Mg-Doped SrMoO3 Perovskites Designed as Anode Materials for Solid Oxide Fuel Cells

PubMed Central

Cascos, Vanessa; Alonso, José Antonio; Fernández-Díaz, María Teresa

2016-01-01

SrMo1−xMxO3−δ (M = Fe and Cr, x = 0.1 and 0.2) oxides have been recently described as excellent anode materials for solid oxide fuel cells at intermediate temperatures (IT-SOFC) with LSGM as the electrolyte. In this work, we have improved their properties by doping with aliovalent Mg ions at the B-site of the parent SrMoO3 perovskite. SrMo1−xMgxO3−δ (x = 0.1, 0.2) oxides have been prepared, characterized and tested as anode materials in single solid-oxide fuel cells, yielding output powers near 900 mW/cm−2 at 850 °C using pure H2 as fuel. We have studied its crystal structure with an “in situ” neutron power diffraction (NPD) experiment at temperatures as high as 800 °C, emulating the working conditions of an SOFC. Adequately high oxygen deficiencies, observed by NPD, together with elevated disk-shaped anisotropic displacement factors suggest a high ionic conductivity at the working temperatures. Furthermore, thermal expansion measurements, chemical compatibility with the LSGM electrolyte, electronic conductivity and reversibility upon cycling in oxidizing-reducing atmospheres have been carried out to find out the correlation between the excellent performance as an anode and the structural features. PMID:28773708

Strontium and Trace Metals in the Mississippi River Mixing Zone

NASA Astrophysics Data System (ADS)

Xu, Y.; Marcantonio, F.

2001-12-01

Strontium is generally believed to be a conservative element, i.e., it is assumed that dissolved Sr moves directly from rivers through estuaries to the ocean. More recently, however, detailed sampling of rivers suggests a weak non-conservative behavior for Sr. Here, we present dissolved and suspended load Sr and trace metal data for samples retrieved along salinity transects in the estuarine mixing zone of the Mississippi River. Our cruises took place during times representing high, falling, and low Mississippi River discharge. Sr concentration and isotopic composition were analyzed for both dissolved particulate loads. Selected particle-reactive or redox-sensitive trace metals (Mn, Fe, U, V, Mo, Ti, and Pb) were analyzed simultaneously. In the dissolved load, Sr showed conservative behavior in both high- and low- discharge periods. Non-conservative behavior of Sr predominated during falling discharge in the summer. Significant positive correlations were found between Sr, Mo and Ti. U and V distributions were found to be essentially controlled by mixing of river water and seawater, but with significantly lower riverine concentrations during high-flow stage. Particulate element concentrations can be quite variable and heterogeneous. In this study, strong correlations were found between particulate Mn (and Fe) concentrations and particulate concentrations of Ti, U, V, and Pb. No such correlations with Mn (or Fe) were found for particulate Sr and Mo. There is a vast hypoxic zone along the coast of Louisiana in the Gulf of Mexico that exists during the summer months. Based on the Sr isotope systematics and the relationships between Sr and trace metals, we believe that this eutrophication may contribute to the non-conservative behaviors of Sr and other trace metals. We discuss the potential implications of this hypothesis on the Sr mass balance of present-day and past seawater.

Characterization of hierarchical α-MoO3 plates toward resistive heating synthesis: electrochemical activity of α-MoO3/Pt modified electrode toward methanol oxidation at neutral pH

NASA Astrophysics Data System (ADS)

Filippo, Emanuela; Baldassarre, Francesca; Tepore, Marco; Guascito, Maria Rachele; Chirizzi, Daniela; Tepore, Antonio

2017-05-01

The growth of MoO3 hierarchical plates was obtained by direct resistive heating of molybdenum foils at ambient pressure in the absence of any catalysts and templates. Plates synthesized after 60 min resistive heating typically grow in an single-crystalline orthorhombic structure that develop preferentially in the [001] direction, and are characterized by high resolution transmission electron microscopy, selected area diffraction pattern and Raman-scattering measurements. They are about 100-200 nm in thickness and a few tens of micrometers in length. As heating time proceeds to 80 min, plates of α-MoO3 form a branched structure. A more attentive look shows that primary plates formed at until 60 min could serve as substrates for the subsequent growth of secondary belts. Moreover, a full electrochemical characterization of α-MoO3 plates on platinum electrodes was done by cyclic voltammetric experiments, at pH 7 in phosphate buffer, to probe the activity of the proposed composite material as anode to methanol electro-oxidation. Reported results indicate that Pt MoO3 modified electrodes are appropriate to develop new an amperometric non-enzymatic sensor for methanol as well as to make anodes suitable to be used in direct methanol fuel cells working at neutral pH.

Encapsulation of Mo2C in MoS2 inorganic fullerene-like nanoparticles and nanotubes

NASA Astrophysics Data System (ADS)

Wiesel, Inna; Popovitz-Biro, Ronit; Tenne, Reshef

2013-01-01

Mo2C nanoparticles encapsulated within MoS2 inorganic fullerene-like nanoparticles and nanotubes were produced by carbothermal reaction at 1200-1300 °C inside a vertical induction furnace. The particles were analyzed using various electron microscopy techniques and complementary methods.

Making the Surface Fleet Green: The DOTMLPF, Policy, and Cost Implications of Using Biofuel in Surface Ships

DTIC Science & Technology

2012-12-01

Navy’s Ships Renewable Fuels Evaluation, 2011) ..25 Table 4. Diesel Injector Component Testing (From U.S. Navy Biofuel Test and Qualification Update...components, including shipboard quality assurance instruments, fuel injector nozzles , fuel nozzle atomization, fuel nozzle fouling, carbon deposition...Leung, Turgeon, & Williams, 2011, p. 7). Table 4 lists the results from component testing conducted on various diesel engine fuel injectors using

Neutron-induced fission cross-section measurement of 234U with quasi-monoenergetic beams in the keV and MeV range using micromegas detectors

NASA Astrophysics Data System (ADS)

Tsinganis, A.; Kokkoris, M.; Vlastou, R.; Kalamara, A.; Stamatopoulos, A.; Kanellakopoulos, A.; Lagoyannis, A.; Axiotis, M.

2017-09-01

Accurate data on neutron-induced fission cross-sections of actinides are essential for the design of advanced nuclear reactors based either on fast neutron spectra or alternative fuel cycles, as well as for the reduction of safety margins of existing and future conventional facilities. The fission cross-section of 234U was measured at incident neutron energies of 560 and 660 keV and 7.5 MeV with a setup based on `microbulk' Micromegas detectors and the same samples previously used for the measurement performed at the CERN n_TOF facility (Karadimos et al., 2014). The 235U fission cross-section was used as reference. The (quasi-)monoenergetic neutron beams were produced via the 7Li(p,n) and the 2H(d,n) reactions at the neutron beam facility of the Institute of Nuclear and Particle Physics at the `Demokritos' National Centre for Scientific Research. A detailed study of the neutron spectra produced in the targets and intercepted by the samples was performed coupling the NeuSDesc and MCNPX codes, taking into account the energy spread, energy loss and angular straggling of the beam ions in the target assemblies, as well as contributions from competing reactions and neutron scattering in the experimental setup. Auxiliary Monte-Carlo simulations were performed with the FLUKA code to study the behaviour of the detectors, focusing particularly on the reproduction of the pulse height spectra of α-particles and fission fragments (using distributions produced with the GEF code) for the evaluation of the detector efficiency. An overview of the developed methodology and preliminary results are presented.

Size distribution of EC, OC and particle-phase PAHs emissions from a diesel engine fueled with three fuels.

PubMed

Lu, Tian; Huang, Zhen; Cheung, C S; Ma, Jing

2012-11-01

The size distribution of elemental carbon (EC), organic carbon (OC) and particle-phase PAHs emission from a direct injection diesel engine fueled with a waste cooking biodiesel, ultra low sulfur diesel (ULSD, 10-ppm-wt), and low sulfur diesel (LSD, 400-ppm-wt) were investigated experimentally. The emission factor of biodiesel EC is 90.6 mg/kh, which decreases by 60.3 and 71.7%, compared with ULSD and LSD respectively and the mass mean diameter (MMD) of EC was also decreased with the use of biodiesel. The effect of biodiesel on OC emission might depend on the engine operation condition, and the difference in OC size distribution is not that significant among the three fuels. For biodiesel, its brake specific emission of particle-phase PAHs is obviously smaller than that from the two diesel fuels, and the reduction effect appears in almost all size ranges. In terms of size distribution, the MMD of PAHs from biodiesel is larger than that from the two diesel fuels, which could be attributed to the more effective reduction on combustion derived PAHs in nuclei mode. The toxicity analysis indicates that biodiesel could reduce the total PAHs emissions, as well as the carcinogenic potency of particle-phase PAHs in almost all the size ranges. Copyright © 2012 Elsevier B.V. All rights reserved.

Associations of Pd, U and Ag in the SiC layer of neutron-irradiated TRISO fuel

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

Lillo, Thomas; Rooyen, Isabella Van

2015-05-01

Knowledge of the associations and composition of fission products in the neutron irradiated SiC layer of high-temperature gas reactor TRISO fuel is important to the understanding of various aspects of fuel performance that presently are not well understood. Recently, advanced characterization techniques have been used to examine fuel particles from the Idaho National Laboratory’s AGR-1 experiment. Nano-sized Ag and Pd precipitates were previously identified in grain boundaries and triple points in the SiC layer of irradiated TRISO nuclear fuel. Continuation of this initial research is reported in this paper and consists of the characterization of a relatively large number ofmore » nano-sized precipitates in three areas of the SiC layer of a single irradiated TRISO nuclear fuel particle using standardless EDS analysis on focused ion beam-prepared transmission electron microscopy samples. Composition and distribution analyses of these precipitates, which were located on grain boundaries, triple junctions and intragranular precipitates, revealed low levels, generally <10 atomic %, of palladium, silver and/or uranium with palladium being the most common element found. Palladium by itself, or associated with either silver or uranium, was found throughout the SiC layer. A small number of precipitates on grain boundaries and triple junctions were found to contain only silver or silver in association with palladium while uranium was always associated with palladium but never found by itself or in association with silver. Intergranular precipitates containing uranium were found to have migrated ~23 μm along a radial direction through the 35 μm thick SiC coating during the AGR-1 experiment while silver-containing intergranular precipitates were found at depths up to ~24 μm in the SiC layer. Also, Pd-rich, nano-precipitates (~10 nm in diameter), without evidence for the presence of either Ag or U, were revealed in intragranular regions throughout the SiC layer. Because not all grain boundaries and triple junctions contained precipitates with fission products and/or uranium, along with the differences in migration behavior between Pd, Ag and U, it was concluded that crystallographic grain boundary and triple junction parameters likely influence migration behavior.« less

Flood-Inundation Maps for the Meramec River at Valley Park and at Fenton, Missouri, 2017

DOT National Transportation Integrated Search

2017-01-01

Two sets of digital flood-inundation map libraries that spanned a combined 16.7-mile reach of the Meramec River that extends upstream from Valley Park, Missouri, to downstream from Fenton, Mo., were created by the U.S. Geological Survey (USGS) in coo...

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

Bi, G.; Liu, C.; Si, S.

This paper was focused on core design, neutronics evaluation and fuel cycle analysis for Thorium-Uranium Breeding Recycle in current PWRs, without any major change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. The fuel cycle analysis indicates that Thorium-Uranium Breeding Recycle is technically feasible in current PWRs. A 4-loop, 193-assembly PWR core utilizing 17 x 17 fuel assemblies (FAs) was taken as the model core. Two mixed cores were investigated respectively loaded with mixed reactor grade Plutonium-Thorium (PuThOX) FAs and mixed reactor grade {sup 233}U-Thorium (U{sub 3}ThOX) FAs on the basis ofmore » reference full Uranium oxide (UOX) equilibrium-cycle core. The UOX/PuThOX mixed core consists of 121 UOX FAs and 72 PuThOX FAs. The reactor grade {sup 233}U extracted from burnt PuThOX fuel was used to fabrication of U{sub 3}ThOX for starting Thorium-. Uranium breeding recycle. In UOX/U{sub 3}ThOX mixed core, the well designed U{sub 3}ThOX FAs with 1.94 w/o fissile uranium (mainly {sup 233}U) were located on the periphery of core as a blanket region. U{sub 3}ThOX FAs remained in-core for 6 cycles with the discharged burnup achieving 28 GWD/tHM. Compared with initially loading, the fissile material inventory in U{sub 3}ThOX fuel has increased by 7% via 1-year cooling after discharge. 157 UOX fuel assemblies were located in the inner of UOX/U{sub 3}ThOX mixed core refueling with 64 FAs at each cycle. The designed UOX/PuThOX and UOX/U{sub 3}ThOX mixed core satisfied related nuclear design criteria. The full core performance analyses have shown that mixed core with PuThOX loading has similar impacts as MOX on several neutronic characteristic parameters, such as reduced differential boron worth, higher critical boron concentration, more negative moderator temperature coefficient, reduced control rod worth, reduced shutdown margin, etc.; while mixed core with U{sub 3}ThOX loading on the periphery of core has no visible impacts on neutronic characteristics compared with reference full UOX core. The fuel cycle analysis has shown that {sup 233}U mono-recycling with U{sub 3}ThOX fuel could save 13% of natural uranium resource compared with UOX once through fuel cycle, slightly more than that of Plutonium single-recycling with MOX fuel. If {sup 233}U multi-recycling with U{sub 3}ThOX fuel is implemented, more natural uranium resource would be saved. (authors)« less

Self-Assembled Nano-energetic Gas Generators based on Bi2O3

NASA Astrophysics Data System (ADS)

Hobosyan, Mkhitar; Trevino, Tyler; Martirosyan, Karen

2012-10-01

Nanoenergetic Gas-Generators are formulations that rapidly release a large amount of gaseous products and generate a fast moving thermal wave. They are mainly based on thermite systems, which are pyrotechnic mixtures of metal powders (fuel- Al, Mg, etc.) and metal oxides (oxidizer, Bi2O3, Fe2O3, WO3, MoO3 etc.) that can generate an exothermic oxidation-reduction reaction referred to as a thermite reaction. A thermite reaction releases a large amount of energy and can generate rapidly extremely high temperatures. The intimate contact between the fuel and oxidizer can be enhanced by use of nano instead of micro particles. The contact area between oxidizer and metal particles depends from method of mixture preparation. In this work we utilize the self-assembly processes, which use the electrostatic forces to produce ordered and self-organized binary systems. In this process the intimate contact significantly enhances and gives the ability to build an energetic material in molecular level, which is crucial for thepressure discharge efficiency of nano-thermites. The DTA-TGA, Zeta-size analysis and FTIR technique were performed to characterize the Bi2O3 particles. The self-assembly of Aluminum and Bi2O3 was conducted in sonic bath with appropriate solvents and linkers. The resultant thermite pressure discharge values were tested in modified Parr reactor. In general, the self-assembled thermites give much higher-pressure discharge values than the thermites prepared with conventional roll-mixing technique.