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Sample records for burnup simulated nitride

  1. Model biases in high-burnup fast reactor simulations

    Energy Technology Data Exchange (ETDEWEB)

    Touran, N.; Cheatham, J.; Petroski, R. [TerraPower LLC, 11235 S.E. 6th St, Bellevue, WA 98004 (United States)

    2012-07-01

    A new code system called the Advanced Reactor Modeling Interface (ARMI) has been developed that loosely couples multiscale, multiphysics nuclear reactor simulations to provide rapid, user-friendly, high-fidelity full systems analysis. Incorporating neutronic, thermal-hydraulic, safety/transient, fuel performance, core mechanical, and economic analyses, ARMI provides 'one-click' assessments of many multi-disciplined performance metrics and constraints that historically require iterations between many diverse experts. The capabilities of ARMI are implemented in this study to quantify neutronic biases of various modeling approximations typically made in fast reactor analysis at an equilibrium condition, after many repetitive shuffles. Sensitivities at equilibrium that result in very high discharge burnup are considered ( and >20% FIMA), as motivated by the development of the Traveling Wave Reactor. Model approximations discussed include homogenization, neutronic and depletion mesh resolution, thermal-hydraulic coupling, explicit control rod insertion, burnup-dependent cross sections, fission product model, burn chain truncation, and dynamic fuel performance. The sensitivities of these approximations on equilibrium discharge burnup, k{sub eff}, power density, delayed neutron fraction, and coolant temperature coefficient are discussed. (authors)

  2. Simulation of triton burn-up in JET plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Loughlin, M.J.; Balet, B.; Jarvis, O.N.; Stubberfield, P.M. [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking

    1994-07-01

    This paper presents the first triton burn-up calculations for JET plasmas using the transport code TRANSP. Four hot ion H-mode deuterium plasmas are studied. For these discharges, the 2.5 MeV emission rises rapidly and then collapses abruptly. This phenomenon is not fully understood but in each case the collapse phase is associated with a large impurity influx known as the ``carbon bloom``. The peak 14 MeV emission occurs at this time, somewhat later than that of the 2.5 MeV neutron peak. The present results give a clear indication that there are no significant departures from classical slowing down and spatial diffusion for tritons in JET plasmas. (authors). 7 refs., 3 figs., 1 tab.

  3. French investigations of high burnup effect on LOCA thermomecanical behavior. Part two. Oxidation and quenching experiments under simulated LOCA conditions with high burnup clad material

    Energy Technology Data Exchange (ETDEWEB)

    GrandJean, C. [IPSN, Cadarache (France); Cauvin, R.; Lebuffe, C. [EDF/SCMI, Chinon (France)] [and others

    1997-01-01

    In the frame of the high burnup fuel studies to support a possible extension of the current discharge burnup limit, experimental programs have been undertaken, jointly by EDF and IPSN in order to study the thermal-shock behavior of high burnup fuel claddings under typical LOCA conditions. The TAGUS program used unirradiated cladding samples, bare or bearing a pre-corrosion state simulating the end-of-life state of high burnup fuel claddings: the TAGCIR program used actually irradiated cladding samples taken from high burnup rods irradiated over 5 cycles in a commercial EDF PWR and having reached a rod burnup close to 60 GWd/tU. The thermal-shock failure tests consisted in oxidizing the cladding samples under steam flow, on both inner and outer faces or on the outer face alone, and subjecting them to a final water quench. The heating was provided by an inductive furnace the power of which being regulated through monitoring of the sample surface temperature with use of a single-wave optical pyrometer. Analysis of the irradiated tests (TAGCIR series) evidenced an increased oxidation rate as compared to similar tests on unirradiated samples. Results of the quenching tests series on unirradiated and irradiated samples are plotted under the usual presentation of failure maps relative to the oxidation parameters ECR (equivalent cladding reacted) or e{sub {beta}} (thickness of the remaining beta phase layer) as a function of the oxidation temperature. Comparison of the failure limits for irradiated specimens to those for unirradiated specimens indicates a lower brittleness under two side oxidation and possibly the opposite under one-side oxidation. The tentative analysis of the oxidation and quenching tests results on irradiated samples reveals the important role played by the hydrogen charged during in-reactor corrosion on the oxidation kinetics and the failure bearing capability of the cladding under LOCA transient conditions.

  4. Computational simulation of fuel burnup estimation for research reactors plate type

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Nadia Rodrigues dos, E-mail: nadiasam@gmail.com [Instituto Federal de Educacao, Ciencia e Tecnologia do Rio de Janeiro (IFRJ), Paracambi, RJ (Brazil); Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes, E-mail: zrlima@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    The aim of this study is to estimate the spatial fuel burnup, through computational simulation, in two research reactors plate type, loaded with dispersion fuel: the benchmark Material Test Research - International Atomic Energy Agency (MTR-IAEA) and a typical multipurpose reactor (MR). The first composed of plates with uranium oxide dispersed in aluminum (UAlx-Al) and a second composed with uranium silicide (U{sub 3}Si{sub 2}) dispersed in aluminum. To develop this work we used the deterministic code, WIMSD-5B, which performs the cell calculation solving the neutron transport equation, and the DF3DQ code, written in FORTRAN, which solves the three-dimensional neutron diffusion equation using the finite difference method. The methodology used was adequate to estimate the spatial fuel burnup , as the results was in accordance with chosen benchmark, given satisfactorily to the proposal presented in this work, even showing the possibility to be applied to other research reactors. For future work are suggested simulations with other WIMS libraries, other settings core and fuel types. Comparisons the WIMSD-5B results with programs often employed in fuel burnup calculations and also others commercial programs, are suggested too. Another proposal is to estimate the fuel burnup, taking into account the thermohydraulics parameters and the Xenon production. (author)

  5. Indium gallium nitride multijunction solar cell simulation using silvaco atlas

    OpenAIRE

    Garcia, Baldomero

    2007-01-01

    This thesis investigates the potential use of wurtzite Indium Gallium Nitride as photovoltaic material. Silvaco Atlas was used to simulate a quad-junction solar cell. Each of the junctions was made up of Indium Gallium Nitride. The band gap of each junction was dependent on the composition percentage of Indium Nitride and Gallium Nitride within Indium Gallium Nitride. The findings of this research show that Indium Gallium Nitride is a promising semiconductor for solar cell use. United...

  6. Burnup simulations of different fuel grades using the MCNPX Monte Carlo code

    Directory of Open Access Journals (Sweden)

    Asah-Opoku Fiifi

    2014-01-01

    Full Text Available Global energy problems range from the increasing cost of fuel to the unequal distribution of energy resources and the potential climate change resulting from the burning of fossil fuels. A sustainable nuclear energy would augment the current world energy supply and serve as a reliable future energy source. This research focuses on Monte Carlo simulations of pressurized water reactor systems. Three different fuel grades - mixed oxide fuel (MOX, uranium oxide fuel (UOX, and commercially enriched uranium or uranium metal (CEU - are used in this simulation and their impact on the effective multiplication factor (Keff and, hence, criticality and total radioactivity of the reactor core after fuel burnup analyzed. The effect of different clad materials on Keff is also studied. Burnup calculation results indicate a buildup of plutonium isotopes in UOX and CEU, as opposed to a decline in plutonium radioisotopes for MOX fuel burnup time. For MOX fuel, a decrease of 31.9% of the fissile plutonium isotope is observed, while for UOX and CEU, fissile plutonium isotopes increased by 82.3% and 83.8%, respectively. Keff results show zircaloy as a much more effective clad material in comparison to zirconium and stainless steel.

  7. Comparison of nuclear data uncertainty propagation methodologies for PWR burn-up simulations

    CERN Document Server

    Diez, Carlos Javier; Hoefer, Axel; Porsch, Dieter; Cabellos, Oscar

    2014-01-01

    Several methodologies using different levels of approximations have been developed for propagating nuclear data uncertainties in nuclear burn-up simulations. Most methods fall into the two broad classes of Monte Carlo approaches, which are exact apart from statistical uncertainties but require additional computation time, and first order perturbation theory approaches, which are efficient for not too large numbers of considered response functions but only applicable for sufficiently small nuclear data uncertainties. Some methods neglect isotopic composition uncertainties induced by the depletion steps of the simulations, others neglect neutron flux uncertainties, and the accuracy of a given approximation is often very hard to quantify. In order to get a better sense of the impact of different approximations, this work aims to compare results obtained based on different approximate methodologies with an exact method, namely the NUDUNA Monte Carlo based approach developed by AREVA GmbH. In addition, the impact ...

  8. MCNPX Monte Carlo burnup simulations of the isotope correlation experiments in the NPP Obrigheim

    Energy Technology Data Exchange (ETDEWEB)

    Cao Yan, E-mail: ycao@anl.go [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Gohar, Yousry [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Broeders, Cornelis H.M. [Forschungszentrum Karlsruhe, Institute for Neutron Physics and Reactor Technology, P.O. Box 3640, 76021 Karlsruhe (Germany)

    2010-10-15

    This paper describes the simulation work of the Isotope Correlation Experiment (ICE) using the MCNPX Monte Carlo computer code package. The Monte Carlo simulation results are compared with the ICE-Experimental measurements for burnup up to 30 GWD/t. The comparison shows the good capabilities of the MCNPX computer code package for predicting the depletion of the uranium fuel and the buildup of the plutonium isotopes in a PWR thermal reactor. The Monte Carlo simulation results show also good agreements with the experimental data for calculating several long-lived and stable fission products. However, for the americium and curium actinides, it is difficult to judge the predication capabilities for these actinides due to the large uncertainties in the ICE-Experimental data. In the MCNPX numerical simulations, a pin cell model is utilized to simulate the fuel lattice of the nuclear power reactor. Temperature dependent libraries based on JEFF3.1 nuclear data files are utilized for the calculations. In addition, temperature dependent libraries based ENDF/B-VII nuclear data files are utilized and the obtained results are very close to the JEFF3.1 results, except for {approx}10% differences in the prediction of the minor actinide isotopes buildup.

  9. Transport and Burnup Numerical Simulation on the Liquid Blanket Burnup of In-Zinerater%In-Zinerater液态包层输运燃耗数值模拟

    Institute of Scientific and Technical Information of China (English)

    师学明; 杨俊云; 刘成安

    2014-01-01

    Z-Pinch惯性约束聚变是未来一种有竞争力的能源候选方案。Z-Pinch驱动的聚变裂变混合堆可高效地嬗变反应堆乏燃料中分离出的超铀元素。对美国Sandia国家实验室提出的In-Zinerater混合堆概念进行了中子学分析和数值模拟。在三维输运燃耗耦合程序MCORGS中增加了处理在线添加燃料与去除裂变产物的功能,实现了对液态燃料燃耗过程的模拟。增加6Li丰度和燃料初装量保持寿期初反应性不变,可以减缓寿期内反应性下降趋势。逐步增加包层内超铀元素装量,可以控制整个寿期内反应性基本恒定。聚变功率取20 MW,通过反应性控制,5年内包层能量放大倍数在160∼180之间,氚增殖比在1.5∼1.7之间,优于In-Zinerater基准设计方案。%Z-Pinch Inertial confinement fusion is a competitive candidate for future energy solution. A fusion-fission hybrid driven by Z-Pinch can be used to transmute transuranic elements from spent fuels of reactors efficiently. Analysis and numerical simulation of blanket neutronics of In-Zinerater, which is a fusion-fission hybrid concept design in Sandia National Laboratories, is given in this paper. Modification to the three dimension transport and burnup code MCORGS are done, so as to simulate continuous feeding and continuous chemical processing of the liquid fuel. Different combination of initial enrichment of 6Li and fuels loading in the blanket are selected to keep the same reactivity at begin of core. By this way, the decreasing trend of reactivity at life of the core can be lowered. The reactivity can be maintained constant by increasing the fuel loading in the core gradually as the burnup deepens. Given a 20 MW fusion power, by reactivity control, the blanket energy multiplication is around 160∼180 and tritium breed ratio 1.5∼1.7 in 5 years, which is a better result than Sandia’s original design.

  10. Estimate of fuel burnup spatial a multipurpose reactor in computer simulation; Estimativa da queima espacial do combustivel de um reator multiproposito por simulacao computacional

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Nadia Rodrigues dos, E-mail: nadia.santos@ifrj.edu.br [Instituto Federal de Educacao, Ciencia e Tecnologia do Rio de Janeiro (IFRJ), Paracambi, RJ (Brazil); Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes, E-mail: malu@ien.gov.br, E-mail: zrlima@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    In previous research, which aimed, through computer simulation, estimate the spatial fuel burnup for the research reactor benchmark, material test research - International Atomic Energy Agency (MTR/IAEA), it was found that the use of the code in FORTRAN language, based on the diffusion theory of neutrons and WIMSD-5B, which makes cell calculation, bespoke be valid to estimate the spatial burnup other nuclear research reactors. That said, this paper aims to present the results of computer simulation to estimate the space fuel burnup of a typical multipurpose reactor, plate type and dispersion. the results were considered satisfactory, being in line with those presented in the literature. for future work is suggested simulations with other core configurations. are also suggested comparisons of WIMSD-5B results with programs often employed in burnup calculations and also test different methods of interpolation values obtained by FORTRAN. Another proposal is to estimate the burning fuel, taking into account the thermohydraulics parameters and the appearance of xenon. (author)

  11. Numerical Simulation of Ballistic Impact of Layered Aluminum Nitride Ceramic

    Science.gov (United States)

    2015-09-01

    configuration simulated in this work duplicates that examined in experiments of Yadav and Ravichandran.1 As shown in Fig. 1a, a WHA (tungsten heavy alloy... WHA ) 157 c aJohnson GR, Holmquist TJ, Beissel SR. Response of aluminum nitride (including a phase change) to large strains, high strain rates, and...results cannot be isolated in the present set of simulations, but possibilities include the following: the WHA material may be weaker than that

  12. Modeling and simulation of bulk gallium nitride power semiconductor devices

    Directory of Open Access Journals (Sweden)

    G. Sabui

    2016-05-01

    Full Text Available Bulk gallium nitride (GaN power semiconductor devices are gaining significant interest in recent years, creating the need for technology computer aided design (TCAD simulation to accurately model and optimize these devices. This paper comprehensively reviews and compares different GaN physical models and model parameters in the literature, and discusses the appropriate selection of these models and parameters for TCAD simulation. 2-D drift-diffusion semi-classical simulation is carried out for 2.6 kV and 3.7 kV bulk GaN vertical PN diodes. The simulated forward current-voltage and reverse breakdown characteristics are in good agreement with the measurement data even over a wide temperature range.

  13. Analysis of high burnup fuel safety issues

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Kim, D. H.; Bang, J. G.; Kim, Y. M.; Yang, Y. S.; Jung, Y. H.; Jeong, Y. H.; Nam, C.; Baik, J. H.; Song, K. W.; Kim, K. S

    2000-12-01

    Safety issues in steady state and transient behavior of high burnup LWR fuel above 50 - 60 MWD/kgU were analyzed. Effects of burnup extension upon fuel performance parameters was reviewed, and validity of both the fuel safety criteria and the performance analysis models which were based upon the lower burnup fuel test results was analyzed. It was found that further tests would be necessary in such areas as fuel failure and dispersion for RIA, and high temperature cladding corrosion and mechanical deformation for LOCA. Since domestic fuels have been irradiated in PWR up to burnup higher than 55 MWD/kgU-rod. avg., it can be said that Korea is in the same situation as the other countries in the high burnup fuel safety issues. Therefore, necessary research areas to be performed in Korea were derived. Considering that post-irradiation examination(PIE) for the domestic fuel of burnup higher than 30 MWD/kgU has not been done so far at all, it is primarily necessary to perform PIE for high burnup fuel, and then simulation tests for RIA and LOCA could be performed by using high burnup fuel specimens. For the areas which can not be performed in Korea, international cooperation will be helpful to obtain the test results. With those data base, safety of high burnup domestic fuels will be confirmed, current fuel safety criteria will be re-evaluated, and finally transient high burnup fuel behavior analysis technology will be developed through the fuel performance analysis code development.

  14. Using Coupled Mesoscale Experiments and Simulations to Investigate High Burn-Up Oxide Fuel Thermal Conductivity

    Science.gov (United States)

    Teague, Melissa C.; Fromm, Bradley S.; Tonks, Michael R.; Field, David P.

    2014-12-01

    Nuclear energy is a mature technology with a small carbon footprint. However, work is needed to make current reactor technology more accident tolerant and to allow reactor fuel to be burned in a reactor for longer periods of time. Optimizing the reactor fuel performance is essentially a materials science problem. The current understanding of fuel microstructure have been limited by the difficulty in studying the structure and chemistry of irradiated fuel samples at the mesoscale. Here, we take advantage of recent advances in experimental capabilities to characterize the microstructure in 3D of irradiated mixed oxide (MOX) fuel taken from two radial positions in the fuel pellet. We also reconstruct these microstructures using Idaho National Laboratory's MARMOT code and calculate the impact of microstructure heterogeneities on the effective thermal conductivity using mesoscale heat conduction simulations. The thermal conductivities of both samples are higher than the bulk MOX thermal conductivity because of the formation of metallic precipitates and because we do not currently consider phonon scattering due to defects smaller than the experimental resolution. We also used the results to investigate the accuracy of simple thermal conductivity approximations and equations to convert 2D thermal conductivities to 3D. It was found that these approximations struggle to predict the complex thermal transport interactions between metal precipitates and voids.

  15. Simulation of nitrogen concentration depth profiles in low temperature nitrided stainless steel

    DEFF Research Database (Denmark)

    Christiansen, Thomas; Dahl, Kristian Vinter; Somers, Marcel A. J.

    2006-01-01

    A numerical model is presented, which simulates nitrogen concentration-depth profiles as obtained with low temperature gaseous nitriding of stainless steel. The evolution of the calculated nitrogen concentration-depth profiles is compared with experimental nitriding kinetics. It is shown that the...

  16. Monte Carlo simulation of terahertz generation in nitrides

    Energy Technology Data Exchange (ETDEWEB)

    Starikov, E.; Shiktorov, P.; Gruzinskis, V. [Semiconductor Physics Institute, Vilnius (Lithuania); Reggiani, L. [Dipartimento di Ingegneria dell' Innovazione, Istituto Nazionale di Fisica della Materia, Universite di Lecce, Lecce (Italy); Varani, L.; Vaissiere, J.C. [Centre d' Electronique et de Micro-Optoelectronique de Montpellier (CNRS UMR 5507), Universite Montpellier II, Montpellier (France); Zhao, Jian H. [SiCLAB, Department of Electrical and Computer Engineering and CAIP Center, Rutgers University, Piscataway, NJ (United States)

    2001-08-13

    The conditions for microwave power generation under the quasi-periodic motion of carriers caused by the combined action of carrier acceleration in a constant electric field and optical phonon emission at low temperatures are analysed by means of Monte Carlo simulations of both small- and large-signal responses in bulk nitrides such as GaN and InN. It is shown that, as a consequence of the high value of the optical phonon energy and the strong electron-phonon interaction, a dynamic negative differential mobility caused by transit-time resonance occurs over a wide frequency range which covers practically the whole submillimetre range and persists in the THz frequency range up to liquid nitrogen temperature. The efficiency of the amplification and generation is found to depend nonmonotonically on: (i) the static and microwave electric field amplitudes, (ii) the generation frequency, and (iii) the carrier concentration. Accordingly, for each generation frequency there exists an optimal range of parameter values. Under optimal conditions we predict a generation efficiency of about 1-2% in the 0.5-1.5 THz frequency range. (author)

  17. Platinum group metal nitrides and carbides: synthesis, properties and simulation

    Energy Technology Data Exchange (ETDEWEB)

    Ivanovskii, Alexander L [Institute of Solid State Chemistry, Urals Branch of the Russian Academy of Sciences, Ekaterinburg (Russian Federation)

    2009-04-30

    Experimental and theoretical data on new compounds, nitrides and carbides of the platinum group 4d and 5d metals (ruthenium, rhodium, palladium, osmium, iridium, platinum), published over the past five years are summarized. The extreme mechanical properties of platinoid nitrides and carbides, i.e., their high strength and low compressibility, are noted. The prospects of further studies and the scope of application of these compounds are discussed.

  18. Experiment on the improvement of sinterability for dry recycling nuclear fuel pellets by using simulated spent PWR fuel of high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Woong Ki; Kim, S. S.; Park, G. I.; Lee, Jae W.; Cho, K. H.; Lee, D. Y.; Lee, Y. S.; Lee, J. W.; Yang, M. S.; Shin, W. C

    2004-09-01

    To study the fabrication characteristics of dry recycling nuclear fuel using spent PWR fuel with high burnup of 60,000 MWd/tU, the fission products of spent PWR fuel was analyzed by ORIGEN-2 code. Simulated spent PWR fuel pellets were fabricated by using UO{sub 2} powder added by the simulated fission products. The simulated dry-recycling-fuel pellets were fabricated by dry recycling fuel fabrication flow including 3 cycle treated OREOX(Oxidation and REduction of OXide fuel) process. A small amount of dopant such as TiO{sub 2}, Nb{sub 2}O{sub 5}, Li{sub 2}O are added to increase sinterability of the OREOX treated powder. As the results of experiments, the densities of sintered pellets without dopant ranged from 10.04 to 10.34 g/cm{sup 3}(94.3 to 97.1% of T.D.), the grain size of the pellets ranged from 3 to 4 {mu}m. The sintered density of the pellets with TiO{sub 2} or Nb{sub 2}O{sub 5} ranged from 10.46 to 10.32 g/cm{sup 3}(98.2 to 96.9 % of T.D.) The grain size of the pellets with TiO{sub 2}, Nb{sub 2}O{sub 5} or Li{sub 2}O ranged from 7.3 to 12.2 {mu}m.

  19. Numerical simulation of discharge plasma generation and nitriding the metals and alloys

    Science.gov (United States)

    Koval, T. V.; Manakov, R. A.; Nguyen Bao, Hung; Tran My, Kim An

    2017-01-01

    This research provides the numerical simulation of the plasma generation in a hollow cathode as well as the diffusion of nitrogen atoms into the metal in the low-pressure glow discharge plasma. The characteristics of the gas discharge were obtained and the relation of the basic technological parameters and the structural and phase state of the nitrided material were defined. Authors provided the comparison of calculations with the experimental results of titanium nitriding by low-pressure glow discharge plasma in a hollow cathode.

  20. Efficient FEM simulation of static and free vibration behavior of single walled boron nitride nanotubes

    Science.gov (United States)

    Giannopoulos, Georgios I.; Kontoni, Denise-Penelope N.; Georgantzinos, Stylianos K.

    2016-08-01

    This paper describes the static and free vibration behavior of single walled boron nitride nanotubes using a structural mechanics based finite element method. First, depending on the type of nanotube under investigation, its three dimensional nanostructure is developed according to the well-known corresponding positions of boron and nitride atoms as well as boron nitride bonds. Then, appropriate point masses are assigned to the atomic positions of the developed space frame. Next, these point masses are suitably interconnected with two-noded, linear, spring-like, finite elements. In order to simulate effectively the interactions observed between boron and nitride atoms within the nanotube, appropriate potential energy functions are introduced for these finite elements. In this manner, various atomistic models for both armchair and zigzag nanotubes with different aspect ratios are numerically analyzed and their effective elastic modulus as well as their natural frequencies and corresponding mode shapes are obtained. Regarding the free vibration analysis, the computed results reveal bending, breathing and axial modes of vibration depending on the nanotube size and chirality as well as the applied boundary support conditions. The longitudinal stiffness of the boron nitride nanotubes is found also sensitive to their geometric characteristics.

  1. Influence of FIMA burnup on actinides concentrations in PWR reactors

    Directory of Open Access Journals (Sweden)

    Oettingen Mikołaj

    2016-01-01

    Full Text Available In the paper we present the study on the dependence of actinides concentrations in the spent nuclear fuel on FIMA burnup. The concentrations of uranium, plutonium, americium and curium isotopes obtained in numerical simulation are compared with the result of the post irradiation assay of two spent fuel samples. The samples were cut from the fuel rod irradiated during two reactor cycles in the Japanese Ohi-2 Pressurized Water Reactor. The performed comparative analysis assesses the reliability of the developed numerical set-up, especially in terms of the system normalization to the measured FIMA burnup. The numerical simulations were preformed using the burnup and radiation transport mode of the Monte Carlo Continuous Energy Burnup Code – MCB, developed at the Department of Nuclear Energy, Faculty of Energy and Fuels of AGH University of Science and Technology.

  2. Corrosion studies with high burnup light water reactor fuel. Release of nuclides into simulated groundwater during accumulated contact time of up to two years

    Energy Technology Data Exchange (ETDEWEB)

    Zwicky, Hans-Urs (Zwicky Consulting GmbH, Remigen (Switzerland)); Low, Jeanett; Ekeroth, Ella (Studsvik Nuclear AB, Nykoeping (Sweden))

    2011-03-15

    In the framework of comprehensive research work supporting the development of a Swedish concept for the disposal of highly radioactive waste and spent fuel, Studsvik has performed a significant number of spent fuel corrosion studies under a variety of different conditions. These experiments, performed between 1990 and 2002, covered a burnup range from 27 to 49 MWd/kgU, which was typical for fuel to be disposed at that time. As part of this work, the so called Series 11 tests were performed under oxidising conditions in synthetic groundwater with fuel samples from a rod irradiated in the Ringhals 1 Boiling Water Reactor (BWR). In the meantime, Swedish utilities tend to increase the discharge burnup of fuel operated in their reactors. This means that knowledge of spent fuel corrosion performance has to be extended to higher burnup as well. Therefore, a series of experiments has been started at Studsvik, aiming at extending the data base acquired in the Series 11 corrosion tests to higher burnup fuel. Fuel burnup leads to complex and significant changes in the composition and properties of the fuel. The transformed microstructure, which is referred to as the high burnup structure or rim structure in the outer region of the fuel, consists of small grains of submicron size and a high concentration of pores of typical diameter 1 to 2 mum. This structure forms in UO{sub 2} fuel at a local burnup above 50 MWd/kgU, as long as the temperature is below 1,000-1,100 deg C. The high burnup at the pellet periphery is the consequence of plutonium build-up by neutron capture in 238U followed by fission of the formed plutonium. The amount of fission products in the fuel increases more or less linearly with burnup, in contrast to alpha emitting actinides that increase above average. As burnup across a spent fuel pellet is not uniform, but increases towards the periphery, the radiation field is also larger at the pellet surface. At the same time, it is easier for water to access the

  3. Modeling of numerical simulation and experimental verification for carburizing-nitriding quenching process

    Institute of Scientific and Technical Information of China (English)

    R. MUKAI; T. MATSUMOTO; JU Dong-ying; T. SUZUKI; H. SAITO; Y. ITO

    2006-01-01

    A model considering quantitative effects of diffused carbon and nitrogen gradients and kinetics of phase transformation is presented to examine metallo-thermo-mechanical behavior during carburized and nitrided quenching. Coupled simulation of diffusion,phase transformation and stress/strain provides the final distribution of carbon and nitrogen contents as well as residual stress and distortion. Effects of both transformation and lattice expansion induced by carbon and nitrogen absorption were introduced into calculating the evolution of the internal stress and strain. In order to verify the method and the results,the simulated distributions of carbon and nitrogen content and residual stress/strain of a ring model during carburized and nitrided quenching were compared with the measured data.

  4. Theoretical simulations of regular and defective aluminium nitride nanotubes

    Energy Technology Data Exchange (ETDEWEB)

    Zhukovskii, Yu F [Institute for Solid State Physics, University of Latvia, Kengaraga 8, Riga LV-1063 (Latvia); Popov, A I [Institute for Solid State Physics, University of Latvia, Kengaraga 8, Riga LV-1063 (Latvia); Balasubramanian, C [Department of Environmental, Occupational and Social Medicine, University of Rome Tor Vergata, Via Montpellier 1, I-00133 Rome (Italy); Bellucci, S [INFN-Laboratori Nazionali di Frascati, Via Enrico Fermi 40, I-00044 Frascati (Italy)

    2007-12-15

    For theoretical simulation on AlN nanotubes (NTs) of different chiralities (armchair-and zigzag-type) and uniform diameters, we have considered their single-walled (SW) 1D periodic models. For this aim, we have performed ab initio DFT calculations on AlN SW NTs using formalism of the localized Gaussian-type atomic functions as implemented in CRYSTAL-03 computer code. We have shown that the smaller the diameter of AlN single-walled nanotube is, the closer its electronic and structural properties to AlN bulk. We have analysed an influence of N vacancies (neutral F centres) created by either soft irradiation of nanotubes or under experimental conditions of their growth, on the atomic and electronic structure of AlN SW NTs. We have found the small inward relaxation of the Al nearest neighbours and the N next-nearest neighbours around each point defect formed on 1 nm AlN NTs of both chiralities. Presence of N vacancy in both types of nanotubes has resulted in appearance of the two defect energy levels in their band gaps consisting of mainly 3s and 3p states of the nearest Al atoms.

  5. Simulation of oxidation-nitridation-induced microstructural degradation in a cracked Ni-based superalloy at high temperature

    Directory of Open Access Journals (Sweden)

    Yuan Kang

    2014-01-01

    Full Text Available In turbine engines, high temperature components made of superalloys may crack in a creep process during service. With the inward flux of the gases, e.g. oxygen and nitrogen, along those cracks, the microstructure of the superalloy substrate nearby the cracks may degrade by internal oxidation and nitridation. The aim of this study is to investigate and simulate the oxidation-nitridation-induced microstructural degradation in superalloys by taking a variant of Ni-based superalloy IN-792 as a sample. After the creep testing of the superalloy in air, the microstructures on the cross section of the superalloy were analysed in a scanning electron microscope, equipped with energy/wavelength dispersive systems. Internal oxidation and nitridation, presenting by Al/Ti oxides and nitrides, were observed under a porous and even cracked Cr-oxide scale which was formed on the superalloy surface or along the creep cracks connecting the superalloy surface. Meanwhile, the reinforcing γ′ precipitates were depleted. Such oxidation-nitridation-induced microstructural degradation was simulated by using an oxidation-diffusion model, focusing the diffusion of the alloying elements in metallic phases of the superalloy.

  6. Simulation of Transport Phenomena in Aluminum Nitride Single-Crystal Growth

    Energy Technology Data Exchange (ETDEWEB)

    de Almeida, V F

    2002-04-03

    The goal of this project is to apply advanced computer-aided modeling techniques for simulating coupled radiation transfer present in the bulk growth of aluminum nitride (AlN) single-crystals. Producing and marketing high-quality single-crystals of AlN is currently the focus of Crystal IS, Inc., which is engaged in building a new generation of substrates for electronic and optical-electronic devices. Modeling and simulation of this company's proprietary innovative processing of AlN can substantially improve the understanding of physical phenomena, assist design, and reduce the cost and time of research activities. This collaborative work supported the goals of Crystal IS, Inc. in process scale-up and fundamental analysis with promising computational tools.

  7. A mathematical model and simulation results of plasma enhanced chemical vapor deposition of silicon nitride films

    Science.gov (United States)

    Konakov, S. A.; Krzhizhanovskaya, V. V.

    2015-01-01

    We developed a mathematical model of Plasma Enhanced Chemical Vapor Deposition (PECVD) of silicon nitride thin films from SiH4-NH3-N2-Ar mixture, an important application in modern materials science. Our multiphysics model describes gas dynamics, chemical physics, plasma physics and electrodynamics. The PECVD technology is inherently multiscale, from macroscale processes in the chemical reactor to atomic-scale surface chemistry. Our macroscale model is based on Navier-Stokes equations for a transient laminar flow of a compressible chemically reacting gas mixture, together with the mass transfer and energy balance equations, Poisson equation for electric potential, electrons and ions balance equations. The chemical kinetics model includes 24 species and 58 reactions: 37 in the gas phase and 21 on the surface. A deposition model consists of three stages: adsorption to the surface, diffusion along the surface and embedding of products into the substrate. A new model has been validated on experimental results obtained with the "Plasmalab System 100" reactor. We present the mathematical model and simulation results investigating the influence of flow rate and source gas proportion on silicon nitride film growth rate and chemical composition.

  8. Molecular dynamics simulation of nano-indentation of (111) cubic boron nitride with optimized Tersoff potential

    Science.gov (United States)

    Zhao, Yinbo; Peng, Xianghe; Fu, Tao; Huang, Cheng; Feng, Chao; Yin, Deqiang; Wang, Zhongchang

    2016-09-01

    We conduct molecular dynamics simulation of nanoindentation on (111) surface of cubic boron nitride and find that shuffle-set dislocations slip along direction on {111} plane at the initial stage of the indentation. The shuffle-set dislocations are then found to meet together, forming surfaces of a tetrahedron. We also find that the surfaces are stacking-fault zones, which intersect with each other, forming edges of stair-rod dislocations along direction. Moreover, we also calculate the generalized stacking fault (GSF) energies along various gliding directions on several planes and find that the GSF energies of the {111} and {111} systems are relatively smaller, indicating that dislocations slip more easily along and directions on the {111} plane.

  9. Calibration of burnup monitor installed in Rokkasho Reprocessing Plant

    Energy Technology Data Exchange (ETDEWEB)

    Oeda, Kaoru; Naito, Hirofumi; Hirota, Masanari [Japan Nuclear Fuel Co. Ltd., Rokkasho, Aomori (Japan); Natsume, Koichiro [Isogo Engineering Center, Toshiba Corporation, Yokohama, Kanagawa (Japan); Kumanomido, Hironori [Nuclear Engineering Laboratory, Toshiba Corporation, Kawasaki, Kanagawa (Japan)

    2000-06-01

    Rokkasho Reprocessing Plant uses burnup credit for criticality control at the Spent Fuel Storage Facility (SFSF) and the Dissolution Facility. A burnup monitor measures nondestructively burnup value of a spent fuel assembly and guarantees the credit for burnup. For practical reasons, a standard radiation source is not used in calibration of the burnup monitor, but the burnup values of many spent fuel assemblies are measured based on operator-declared burnup values. This paper describes the concept of burnup credit, the burnup monitor, and the calibration method. It is concluded, from the results of calibration tests, that the calibration method is valid. (author)

  10. Tribological, electrochemical and tribo-electrochemical characterization of bare and nitrided Ti6Al4V in simulated body fluid solution

    Energy Technology Data Exchange (ETDEWEB)

    Manhabosco, T.M., E-mail: tmanhabosco@yahoo.com.b [Physics Departament, Federal University of Ouro Preto, Campus Universitario Morro do Cruzeiro/ICEBS/DEFIS/35400-000, Ouro Preto, Minas Gerais (Brazil); Tamborim, S.M. [Metallurgy Department, Laboratory of Corrosion Research, Federal University of Rio Grande do Sul, Av. Bento Goncalves 9500/75/232, 91501-970 Porto Alegre, Rio Grande do Sul (Brazil); Santos, C.B. dos [Fraunhofer-Institut/IPA Institut fuer Produktionstechnik und Automatisierung, Nobelstrasse 12, Sttutgart 70569 (Germany); Mueller, I.L., E-mail: ilmuller@ufrgs.b [Metallurgy Department, Laboratory of Corrosion Research, Federal University of Rio Grande do Sul, Av. Bento Goncalves 9500/75/232, 91501-970 Porto Alegre, Rio Grande do Sul (Brazil)

    2011-05-15

    Research highlights: {yields} Tribocorrosion of bare and nitrided Ti6Al4V in simulated body fluid is studied. {yields} The alloy presents great tendency to repassivate when its oxide is damaged by wear. {yields} Nitriding increases Ti6Al4V resistance to wear-corrosion at open circuit potential. {yields} EIS results confirm the improved anticorrosion properties of the nitride layer. {yields} Anodic potentials (+0.4V{sub SCE}) impair tribocorrosion resistance of the alloy. - Abstract: Tribological, electrochemical and tribo-electrochemical behaviour of bare and nitrided Ti6Al4V alloy was studied. Scanning Electron Microscopy (SEM), X-ray diffraction and microhardness profile were used to characterize the nitrided Ti6Al4V. The anticorrosive properties of nitrided Ti6Al4V in phosphate buffer saline solution (PBS), simulating the body environment, were evaluated by Electrochemical Impedance Spectroscopy (EIS). Nitriding increased the alloy resistance to corrosion and to dry wear. Resistance to tribocorrosion in PBS at the open circuit potential (OCP) for the nitrided alloy was also significantly increased compared to the bare alloy; nevertheless at an anodic potential this influence became less important.

  11. The vibration properties of the (n,0) boron nitride nanotubes from ab initio quantum chemical simulations.

    Science.gov (United States)

    Erba, A; Ferrabone, M; Baima, J; Orlando, R; Rérat, M; Dovesi, R

    2013-02-07

    The vibration spectrum of single-walled zigzag boron nitride (BN) nanotubes is simulated with an ab initio periodic quantum chemical method. The trend towards the hexagonal monolayer (h-BN) in the limit of large tube radius R is explored for a variety of properties related to the vibrational spectrum: vibration frequencies, infrared intensities, oscillator strengths, and vibration contributions to the polarizability tensor. The (n,0) family is investigated in the range from n = 6 (24 atoms in the unit cell and tube radius R = 2.5 Å) to n = 60 (240 atoms in the cell and R = 24.0 Å). Simulations are performed using the CRYSTAL program which fully exploits the rich symmetry of this class of one-dimensional periodic systems: 4n symmetry operators for the general (n,0) tube. Three sets of infrared active phonon bands are found in the spectrum. The first one lies in the 0-600 cm(-1) range and goes regularly to zero when R increases; the connection between these normal modes and the elastic and piezoelectric constants of h-BN is discussed. The second (600-800 cm(-1)) and third (1300-1600 cm(-1)) sets tend regularly, but with quite different speed, to the optical modes of the h-BN layer. The vibrational contribution of these modes to the two components (parallel and perpendicular) of the polarizability tensor is also discussed.

  12. Fission-gas release at extended burnups: effect of two-dimensional heat transfer

    Energy Technology Data Exchange (ETDEWEB)

    Tayal, M. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Yu, S.D. [Ryerson Polytechnic Univ., Toronto, Ontario (Canada); Lau, J.H.K

    2000-09-01

    To better simulate the performance of high-burnup CANDU fuel, a two-dimensional model for heat transfer between the pellet and the sheath has been added to the computer code ELESTRES. The model covers four relative orientations of the pellet and the sheath and their impacts on heat transfer and fission-gas release. The predictions of the code were compared to a database of 27 experimental irradiations involving extended burnups and normal burnups. The calculated values of fission gas release matched the measurements to an average of 94%. Thus, the two-dimensional heat transfer model increases the versatility of the ELESTRES code to better simulate fuels at normal as well as at extended burnups. (author)

  13. OREST - The hammer-origen burnup program system

    Energy Technology Data Exchange (ETDEWEB)

    Hesse, U. (Gesellschaft fur Reaktorsicherheit mbH Forschungsgelande, 8046 Garching bei Munchen (DE))

    1988-08-01

    Reliable prediction of the characteristics of irradiated light water reactor fuels (e.g., afterheat power, neutron and gamma radiation sources, final uranium and plutonium contents) is needed for many aspects of the nuclear fuel cycle. Two main problems must be solved: the simulation of all isotopic nuclear reactions and the simulation of neutron fluxes setting the reactions in motion. In state-of-the-art computer techniques, a combination of specialized codes for lattice cell and burnup calculations is preferred to solve these cross-linked problems in time or burnup step approximation. In the program system OREST, developed for official and commercial tasks in the Federal Republic of Germany nuclear fuel cycle, the well-known codes HAMMER and ORIGEN and directly coupled with a fuel rod temperature module.

  14. Electronic properties of group III-A nitride sheets by molecular simulation

    Energy Technology Data Exchange (ETDEWEB)

    Anota, Ernesto Chigo; Cocoletzi, Heriberto Hernandez [Facultad de Ingenieria Quimica, BUAP, Avenida San Claudio y 18 sur S/N Edificio 106A CU San Manuel Puebla, 72570 (Mexico); Villanueva, Martin Salazar [Facultad de Ingenieria, BUAP, Puebla (Mexico)

    2010-07-15

    We have performed first principles total energy calculations to investigate the structural and reactivity parameters of novel N{sub 12}X{sub 12}H{sub 12} (X=B, Al, Ga, In, Tl) nitrides, in their coronene-like (C{sub 24}H{sub 12}) structure to simulate these sheets. The exchange and correlations potential energies are treated in the Generalized Gradient Approximation (GGA), and the Local Density Approximation (LDA) within the parameterization of Perdew-Wang and Perdew-Burke-Ernzerhof (PWC, PBE) and the double Numeric plus polarization (DNP) atomic base. The chemical potential, hardness and electrophilicity index, as well as bond length are reported. The bond length of the structures is similar to the bulk. The gap between the HOMO and LUMO decreases from BN (5.18 eV) to TlN (1.76 eV). At the same time, the polarity increases except for TlN. (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  15. Simulated radiation effects in the superinsulating phase of titanium nitride films

    Directory of Open Access Journals (Sweden)

    Vujisić Miloš Lj.

    2011-01-01

    Full Text Available This paper investigates possible effects of alpha particle and ion beam irradiation on the properties of the superinsulating phase, recently observed in titanium nitride films, by using numerical simulation of particle transport. Unique physical properties of the superinsulating state are considered by relying on a two-dimensional Josephson junction array as a model of material structure. It is suggested that radiation-induced change of the Josephson junction charging energy would not affect the current-voltage characteristics of the superinsulating film significantly. However, it is theorized that a relapse to an insulating state with thermally activated resistance is possible, due to radiation-induced disruption of the fine-tuned granular structure. The breaking of Cooper pairs caused by incident and displaced ions may also destroy the conditions for a superinsulating phase to exist. Finally, even the energy loss to phonons can influence the superinsulating state, by increasing the effective temperature of the phonon thermostat, thereby reestablishing means for an energy exchange that can support Cooper pair tunneling.

  16. Encapsulation of cisplatin as an anti-cancer drug into boron-nitride and carbon nanotubes: Molecular simulation and free energy calculation

    Energy Technology Data Exchange (ETDEWEB)

    Roosta, Sara [Molecular Simulation Research Laboratory, Department of Chemistry, Iran University of Science & Technology, Tehran (Iran, Islamic Republic of); Hashemianzadeh, Seyed Majid, E-mail: hashemianzadeh@iust.ac.ir [Molecular Simulation Research Laboratory, Department of Chemistry, Iran University of Science & Technology, Tehran (Iran, Islamic Republic of); Ketabi, Sepideh, E-mail: sepidehketabi@yahoo.com [Department of Chemistry, East Tehran Branch, Islamic Azad University, Tehran (Iran, Islamic Republic of)

    2016-10-01

    Encapsulation of cisplatin anticancer drug into the single walled (10, 0) carbon nanotube and (10, 0) boron-nitride nanotube was investigated by quantum mechanical calculations and Monte Carlo Simulation in aqueous solution. Solvation free energies and complexation free energies of the cisplatin@ carbon nanotube and cisplatin@ boron-nitride nanotube complexes was determined as well as radial distribution functions of entitled compounds. Solvation free energies of cisplatin@ carbon nanotube and cisplatin@ boron-nitride nanotube were − 4.128 kcal mol{sup −1} and − 2457.124 kcal mol{sup −1} respectively. The results showed that cisplatin@ boron-nitride nanotube was more soluble species in water. In addition electrostatic contribution of the interaction of boron- nitride nanotube complex and solvent was − 281.937 kcal mol{sup −1} which really more than Van der Waals and so the electrostatic interactions play a distinctive role in the solvation free energies of boron- nitride nanotube compounds. On the other hand electrostatic part of the interaction of carbon nanotube complex and solvent were almost the same as Van der Waals contribution. Complexation free energies were also computed to study the stability of related structures and the free energies were negative (− 374.082 and − 245.766 kcal mol{sup −1}) which confirmed encapsulation of drug into abovementioned nanotubes. However, boron-nitride nanotubes were more appropriate for encapsulation due to their larger solubility in aqueous solution. - Highlights: • Solubility of cisplatin@ boron-nitride nanotube is larger than cisplatin@ carbon nanotube. • Boron- nitride nanotube complexes have larger electrostatic contribution in solvation free energy. • Complexation free energies confirm encapsulation of drug into the nanotubes in aqueous solution. • Boron- nitride nanotubes are appropriate drug delivery systems compared with carbon nanotubes.

  17. A high burnup model developed for the DIONISIO code

    Energy Technology Data Exchange (ETDEWEB)

    Soba, A. [U.A. Combustibles Nucleares, Comisión Nacional de Energía Atómica, Avenida del Libertador 8250, 1429 Buenos Aires (Argentina); Denis, A., E-mail: denis@cnea.gov.ar [U.A. Combustibles Nucleares, Comisión Nacional de Energía Atómica, Avenida del Libertador 8250, 1429 Buenos Aires (Argentina); Romero, L. [U.A. Reactores Nucleares, Comisión Nacional de Energía Atómica, Avenida del Libertador 8250, 1429 Buenos Aires (Argentina); Villarino, E.; Sardella, F. [Departamento Ingeniería Nuclear, INVAP SE, Comandante Luis Piedra Buena 4950, 8430 San Carlos de Bariloche, Río Negro (Argentina)

    2013-02-15

    A group of subroutines, designed to extend the application range of the fuel performance code DIONISIO to high burn up, has recently been included in the code. The new calculation tools, which are tuned for UO{sub 2} fuels in LWR conditions, predict the radial distribution of power density, burnup, and concentration of diverse nuclides within the pellet. The balance equations of all the isotopes involved in the fission process are solved in a simplified manner, and the one-group effective cross sections of all of them are obtained as functions of the radial position in the pellet, burnup, and enrichment in {sup 235}U. In this work, the subroutines are described and the results of the simulations performed with DIONISIO are presented. The good agreement with the data provided in the FUMEX II/III NEA data bank can be easily recognized.

  18. A high burnup model developed for the DIONISIO code

    Science.gov (United States)

    Soba, A.; Denis, A.; Romero, L.; Villarino, E.; Sardella, F.

    2013-02-01

    A group of subroutines, designed to extend the application range of the fuel performance code DIONISIO to high burn up, has recently been included in the code. The new calculation tools, which are tuned for UO2 fuels in LWR conditions, predict the radial distribution of power density, burnup, and concentration of diverse nuclides within the pellet. The balance equations of all the isotopes involved in the fission process are solved in a simplified manner, and the one-group effective cross sections of all of them are obtained as functions of the radial position in the pellet, burnup, and enrichment in 235U. In this work, the subroutines are described and the results of the simulations performed with DIONISIO are presented. The good agreement with the data provided in the FUMEX II/III NEA data bank can be easily recognized.

  19. Numerical investigation of sputtering power effect on nano-tribological properties of tantalum-nitride film using molecular dynamics simulation

    Science.gov (United States)

    Firouzabadi, S. S.; Dehghani, K.; Naderi, M.; Mahboubi, F.

    2016-03-01

    In the present work, surface profile of tantalum-nitride films, deposited with different sputtering power is analyzed using atomic force microscopy (AFM) in order to find out the effect of sputtering power density on the surface properties. In this regard, micron size tantalum nitride films were deposited using reactive magnetron sputtering system with sputtering power density of 1.5-3 W/cm2 in argon environment mixed with nitrogen. The process was then simulated numerically using molecular dynamics simulation (MD) and Morse interatomic potential in order to study the mechanism of surface roughness variation with sputtering power. According to experimental results, with increasing the sputtering power density from 1.5 to 3, the surface roughness decreased at first followed by an increase. The reason of this behavior was investigated using MD simulation. In this regard, the effect of sputtering power was attributed into two different phenomena: (i) the deposition rate and (ii) the incident-atom energy. Simulation was performed at different deposition rates from 33 to 666 atoms/ps and different sputtered atom energies from .31 to 7.67 eV. It is found that increasing the incident-atom energy causes the atoms to move along the surface and reduces the surface roughness. In contrast, increasing the deposition rate restricted the surface mobility of atoms and roughened the surface. It is also indicated that an increase in the sputtering power cannot increase the adatoms kinetic energy continuously and so the surface diffusion of atoms. Therefore, there is an optimum value for minimum roughness of surface in sputter deposition of tantalum-nitride films which is in agreement with simulation findings.

  20. Encapsulation of cisplatin as an anti-cancer drug into boron-nitride and carbon nanotubes: Molecular simulation and free energy calculation.

    Science.gov (United States)

    Roosta, Sara; Hashemianzadeh, Seyed Majid; Ketabi, Sepideh

    2016-10-01

    Encapsulation of cisplatin anticancer drug into the single walled (10, 0) carbon nanotube and (10, 0) boron-nitride nanotube was investigated by quantum mechanical calculations and Monte Carlo Simulation in aqueous solution. Solvation free energies and complexation free energies of the cisplatin@ carbon nanotube and cisplatin@ boron-nitride nanotube complexes was determined as well as radial distribution functions of entitled compounds. Solvation free energies of cisplatin@ carbon nanotube and cisplatin@ boron-nitride nanotube were -4.128kcalmol(-1) and -2457.124kcalmol(-1) respectively. The results showed that cisplatin@ boron-nitride nanotube was more soluble species in water. In addition electrostatic contribution of the interaction of boron- nitride nanotube complex and solvent was -281.937kcalmol(-1) which really more than Van der Waals and so the electrostatic interactions play a distinctive role in the solvation free energies of boron- nitride nanotube compounds. On the other hand electrostatic part of the interaction of carbon nanotube complex and solvent were almost the same as Van der Waals contribution. Complexation free energies were also computed to study the stability of related structures and the free energies were negative (-374.082 and -245.766kcalmol(-1)) which confirmed encapsulation of drug into abovementioned nanotubes. However, boron-nitride nanotubes were more appropriate for encapsulation due to their larger solubility in aqueous solution.

  1. New burnup calculation of TRIGA IPR-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Meireles, Sincler P. de; Campolina, Daniel de A.M.; Santos, Andre A. Campagnole dos; Menezes, Maria A.B.C.; Mesquita, Amir Z., E-mail: sinclercdtn@hotmail.com.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil, operates since 1960.The reactor is operating for more than fifty years and has a long history of operation. Determining the current composition of the fuel is very important to calculate various parameters. The reactor burnup calculation has been performed before, however, new techniques, methods, software and increase of the processing capacity of the new computers motivates new investigations to be performed. This work presents the evolution of effective multiplication constant and the results of burnup. This new model has a more detailed geometry with the introduction of the new devices, like the control rods and the samarium discs. This increase of materials in the simulation in burnup calculation was very important for results. For these series of simulations a more recently cross section library, ENDF/B-VII, was used. To perform the calculations two Monte Carlo particle transport code were used: Serpent and MCNPX. The results obtained from two codes are presented and compared with previous studies in the literature. (author)

  2. Fuel burnup calculation of a research reactor plate element

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Nadia Rodrigues dos; Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes, E-mail: nadiasam@gmail.com, E-mail: zrlima@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    This work consists in simulating the burnup of two different plate type fuel elements, where one is the benchmark MTR of the IAEA, which is made of an alloy of uranium and aluminum, while the other belonging to a typical multipurpose reactor is composed of an alloy of uranium and silicon. The simulation is performed using the WIMSD-5B computer code, which makes use of deterministic methods for solving neutron transport. In developing this task, fuel element equivalent cells were calculated representing each of the reactors to obtain the initial concentrations of each isotope constituent element of the fuel cell and the thicknesses corresponding to each region of the cell, since this information is part of the input data. The compared values of the k∞ showed a similar behavior for the case of the MTR calculated with the WIMSD-5B and EPRI-CELL codes. Relating the graphs of the concentrations in the burnup of both reactors, there are aspects very similar to each isotope selected. The application WIMSD-5B code to calculate isotopic concentrations and burnup of the fuel element, proved to be satisfactory for the fulfillment of the objective of this work. (author)

  3. Thermodynamic Studies of Decane on Boron Nitride and Graphite Substrates Using Synchrotron Radiation and Molecular Dynamics Simulations

    Science.gov (United States)

    Strange, Nicholas; Arnold, Thomas; Forster, Matthew; Parker, Julia; Larese, J. Z.; Diamond Light Source Collaboration; University of Tennessee Team

    2014-03-01

    Hexagonal boron nitride (hBN) has a lattice structure similar to that of graphite with a slightly larger lattice parameter in the basal plane. This, among other properties, makes it an excellent substrate in place of graphite, eliciting some important differences. This work is part of a larger effort to examine the interaction of alkanes with magnesium oxide, graphite, and boron nitride surfaces. In our current presentation, we will discuss the interaction of decane with these surfaces. Decane exhibits a fully commensurate structure on graphite and hBN at monolayer coverages. In this particular experiment, we have examined the monolayer structure of decane adsorbed on the basal plane of hBN using synchrotron x-ray radiation at Diamond Light Source. Additionally, we have examined the system experimentally with volumetric isotherms as well as computationally using molecular dynamics simulations. The volumetric isotherms allow us to calculate properties which provide important information about the adsorbate's interaction with not only neighboring molecules, but also the interaction with the adsorbent boron nitride.

  4. Investigation on using neutron counting techniques for online burnup monitoring of pebble bed reactor fuels

    Science.gov (United States)

    Zhao, Zhongxiang

    Modular Pebble Bed Reactor (MPBR) is a high temperature gas-cooled nuclear power reactor. This project investigated the feasibility of using the passive neutron counting and active neutron/gamma counting for the on line fuel burnup measurement for MPBR. To investigate whether there is a correlation between neutron emission and fuel burnup, the MPBR fuel depletion was simulated under different irradiation conditions by ORIGEN2. It was found that the neutron emission from an irradiated pebble increases with burnup super-linearly and reaches to 104 neutron/sec/pebble at the discharge burnup. The photon emission from an irradiated pebble was found to be in the order of 1013 photon/sec/pebble at all burnup levels. Analysis shows that the neutron emission rate of an irradiated pebble is sensitive to its burnup history and the spectral-averaged one-group cross sections used in the depletion calculations, which consequently leads to large uncertainty in the correlation between neutron emission and burnup. At low burnup levels, the uncertainty in the neutron emission/burnup correlation is too high and the neutron emission rate is too low so that it is impossible to determine a pebble's burnup by on-line neutron counting at low burnup levels. At high burnup levels, the uncertainty in the neutron emission rate becomes less but is still large in quantity. However, considering the super-linear feature of the correlation, the uncertainty in burnup determination was found to be ˜7% at the discharge burnup, which is acceptable. Therefore, total neutron emission rate of a pebble can be used as a burnup indicator to determine whether a pebble should be discharged or not. The feasibility of using passive neutron counting methods for the on-line burnup measurement was investigated by using a general Monte Carlo code, MCNP, to assess the detectability of the neutron emission and the capability to discriminate gamma noise by commonly used neutron detectors. It was found that both He-3

  5. Dislocation Emission at the Silicon/Silicon Nitride Interface: A Million Atom Molecular Dynamics Simulation on Parallel Computers

    Science.gov (United States)

    Bachlechner, Martina E.; Omeltchenko, Andrey; Nakano, Aiichiro; Kalia, Rajiv K.; Vashishta, Priya; Ebbsjö, Ingvar; Madhukar, Anupam

    2000-01-01

    Mechanical behavior of the Si\\(111\\)/Si3N4\\(0001\\) interface is studied using million atom molecular dynamics simulations. At a critical value of applied strain parallel to the interface, a crack forms on the silicon nitride surface and moves toward the interface. The crack does not propagate into the silicon substrate; instead, dislocations are emitted when the crack reaches the interface. The dislocation loop propagates in the \\(1¯ 1¯1\\) plane of the silicon substrate with a speed of 500 \\(+/-100\\) m/s. Time evolution of the dislocation emission and nature of defects is studied.

  6. Dislocation Emission at the Silicon/Silicon Nitride Interface: A Million Atom Molecular Dynamics Simulation on Parallel Computers

    Energy Technology Data Exchange (ETDEWEB)

    Bachlechner, Martina E.; Omeltchenko, Andrey; Nakano, Aiichiro; Kalia, Rajiv K.; Vashishta, Priya; Ebbsjoe, Ingvar; Madhukar, Anupam

    2000-01-10

    Mechanical behavior of the Si(111)/Si{sub 3}N{sub 4} (0001) interface is studied using million atom molecular dynamics simulations. At a critical value of applied strain parallel to the interface, a crack forms on the silicon nitride surface and moves toward the interface. The crack does not propagate into the silicon substrate; instead, dislocations are emitted when the crack reaches the interface. The dislocation loop propagates in the (1 11) plane of the silicon substrate with a speed of 500 ({+-}100) m/s . Time evolution of the dislocation emission and nature of defects is studied. (c) 2000 The American Physical Society.

  7. A mathematical model and simulation results of plasma enhanced chemical vapor deposition of silicon nitride films

    NARCIS (Netherlands)

    Konakov, S.A.; Krzhizhanovskaya, V.V.

    2015-01-01

    We developed a mathematical model of Plasma Enhanced Chemical Vapor Deposition (PECVD) of silicon nitride thin films from SiH4-NH3-N2-Ar mixture, an important application in modern materials science. Our multiphysics model describes gas dynamics, chemical physics, plasma physics and electrodynamics.

  8. Integrated burnup calculation code system SWAT

    Energy Technology Data Exchange (ETDEWEB)

    Suyama, Kenya; Hirakawa, Naohiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Iwasaki, Tomohiko

    1997-11-01

    SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. It enables us to analyze the burnup problem using neutron spectrum depending on environment of irradiation, combining SRAC which is Japanese standard thermal reactor analysis code system and ORIGEN2 which is burnup code widely used all over the world. SWAT makes effective cross section library based on results by SRAC, and performs the burnup analysis with ORIGEN2 using that library. SRAC and ORIGEN2 can be called as external module. SWAT has original cross section library on based JENDL-3.2 and libraries of fission yield and decay data prepared from JNDC FP Library second version. Using these libraries, user can use latest data in the calculation of SWAT besides the effective cross section prepared by SRAC. Also, User can make original ORIGEN2 library using the output file of SWAT. This report presents concept and user`s manual of SWAT. (author)

  9. Extended calculations of OECD/NEA phase II-C burnup credit criticality benchmark problem for PWR spent fuel transport cask by using MCNP-4B2 code and JENDL-3.2 library

    Energy Technology Data Exchange (ETDEWEB)

    Kuroishi, Takeshi; Hoang, Anh Tuan; Nomura, Yasushi; Okuno, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    The reactivity effect of the asymmetry of axial burnup profile in burnup credit criticality safety is studied for a realistic PWR spent fuel transport cask proposed in the current OECD/NEA Phase II-C benchmark problem. The axial burnup profiles are simulated in 21 material zones based on in-core flux measurements varying from strong asymmetry to more or less no asymmetry. Criticality calculations in a 3-D model have been performed using the continuous energy Monte Carlo code MCNP-4B2 and the nuclear data library JENDL-3.2. Calculation conditions are determined with consideration of the axial fission source convergence. Calculations are carried out not only for cases proposed in the benchmark but also for additional cases assuming symmetric burnup profile. The actinide-only approach supposed for first domestic introduction of burnup credit into criticality evaluation is also considered in addition to the actinide plus fission product approach adopted in the benchmark. The calculated results show that k{sub eff} and the end effect increase almost linearly with increasing burnup axial offset that is defined as one of typical parameters showing the intensity of axial burnup asymmetry. The end effect is more sensitive to the asymmetry of burnup profile for the higher burnup. For an axially distributed burnup, the axial fission source distribution becomes strongly asymmetric as its peak shifts toward the top end of the fuel's active zone where the local burnup is less than that of the bottom end. The peak of fission source distribution becomes higher with the increase of either the asymmetry of burnup profile or the assembly-averaged burnup. The conservatism of the assumption of uniform axial burnup based on the actinide-only approach is estimated quantitatively in comparison with the k{sub eff} result calculated with experiment-based strongest asymmetric axial burnup profile with the actinide plus fission product approach. (author)

  10. Extended calculations of OECD/NEA phase II-C burnup credit criticality benchmark problem for PWR spent fuel transport cask by using MCNP-4B2 code and JENDL-3.2 library

    Energy Technology Data Exchange (ETDEWEB)

    Kuroishi, Takeshi; Hoang, Anh Tuan; Nomura, Yasushi; Okuno, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    The reactivity effect of the asymmetry of axial burnup profile in burnup credit criticality safety is studied for a realistic PWR spent fuel transport cask proposed in the current OECD/NEA Phase II-C benchmark problem. The axial burnup profiles are simulated in 21 material zones based on in-core flux measurements varying from strong asymmetry to more or less no asymmetry. Criticality calculations in a 3-D model have been performed using the continuous energy Monte Carlo code MCNP-4B2 and the nuclear data library JENDL-3.2. Calculation conditions are determined with consideration of the axial fission source convergence. Calculations are carried out not only for cases proposed in the benchmark but also for additional cases assuming symmetric burnup profile. The actinide-only approach supposed for first domestic introduction of burnup credit into criticality evaluation is also considered in addition to the actinide plus fission product approach adopted in the benchmark. The calculated results show that k{sub eff} and the end effect increase almost linearly with increasing burnup axial offset that is defined as one of typical parameters showing the intensity of axial burnup asymmetry. The end effect is more sensitive to the asymmetry of burnup profile for the higher burnup. For an axially distributed burnup, the axial fission source distribution becomes strongly asymmetric as its peak shifts toward the top end of the fuel's active zone where the local burnup is less than that of the bottom end. The peak of fission source distribution becomes higher with the increase of either the asymmetry of burnup profile or the assembly-averaged burnup. The conservatism of the assumption of uniform axial burnup based on the actinide-only approach is estimated quantitatively in comparison with the k{sub eff} result calculated with experiment-based strongest asymmetric axial burnup profile with the actinide plus fission product approach. (author)

  11. Thermo-piezo-electro-mechanical simulation of AlGaN (aluminum gallium nitride) / GaN (gallium nitride) High Electron Mobility Transistors

    Science.gov (United States)

    Stevens, Lorin E.

    Due to the current public demand of faster, more powerful, and more reliable electronic devices, research is prolific these days in the area of high electron mobility transistor (HEMT) devices. This is because of their usefulness in RF (radio frequency) and microwave power amplifier applications including microwave vacuum tubes, cellular and personal communications services, and widespread broadband access. Although electrical transistor research has been ongoing since its inception in 1947, the transistor itself continues to evolve and improve much in part because of the many driven researchers and scientists throughout the world who are pushing the limits of what modern electronic devices can do. The purpose of the research outlined in this paper was to better understand the mechanical stresses and strains that are present in a hybrid AlGaN (Aluminum Gallium Nitride) / GaN (Gallium Nitride) HEMT, while under electrically-active conditions. One of the main issues currently being researched in these devices is their reliability, or their consistent ability to function properly, when subjected to high-power conditions. The researchers of this mechanical study have performed a static (i.e. frequency-independent) reliability analysis using powerful multiphysics computer modeling/simulation to get a better idea of what can cause failure in these devices. Because HEMT transistors are so small (micro/nano-sized), obtaining experimental measurements of stresses and strains during the active operation of these devices is extremely challenging. Physical mechanisms that cause stress/strain in these structures include thermo-structural phenomena due to mismatch in both coefficient of thermal expansion (CTE) and mechanical stiffness between different materials, as well as stress/strain caused by "piezoelectric" effects (i.e. mechanical deformation caused by an electric field, and conversely voltage induced by mechanical stress) in the AlGaN and GaN device portions (both

  12. Simulation of STM technique for electron transport through boron-nitride nanotubes

    Energy Technology Data Exchange (ETDEWEB)

    Ganji, M.D. [Department of Chemistry, Azad University, Ghaemshahr Branch, Mazandaran (Iran, Islamic Republic of)], E-mail: ganji_md@yahoo.com; Mohammadi-nejad, A. [Department of Computer Engineering, Olom Fonon University of Mazandaran, Babol (Iran, Islamic Republic of)

    2008-06-30

    We report first-principles calculations on the electrical transport properties of boron-nitrid nanotubes (BNNTs). We consider a single walled (5,0) boron-nitrid nanotube sandwiched between an Au(1 0 0) substrate and a monatomic Au scanning tunneling microscope (STM) tip. Lateral motion of the tip over the nanotube wall cause it to change from one conformation class to the others and to switch between a strongly and a weakly conducting state. Thus, surprisingly, despite their apparent simplicity these Au/BNNT/Au nanowires are shown to be a convenient switch. Experiments with a conventional STM are proposed to test these predictions. The projection of the density of states (PDOS) and the transmission coefficients T(E) of the two-probe systems at zero bias are analyzed, and it suggests that the variation of the coupling between the wire and the electrodes leads to switching behaviour.

  13. Simulation of mechanical properties and residual stress of nanostructural coatings based on transition metals nitrides

    Science.gov (United States)

    Danilyuk, Alexander L.; Shaposhnikov, Victor L.; Filonov, Andrew B.; Anischik, Victor M.; Uglov, Vladimir V.; Kuleshov, Andrew K.; Danilyuk, Maxim A.

    2008-07-01

    Physical properties of novel nanostructural coatings, formed by ion-plasmous flux from solid solutions of transition and refractory metals (Ti, Zr, Cr) have been intensively studied to enhance characteristics of tool materials. We have developed the modeling technique for effective predictions of internal stresses and calculation of elastic properties of nanostructural coatings composed of metal nitrides. Quantum-mechanical modeling of microstructure, elastic constants, bulk modulus and residual stress for binary and ternary metal nitride clusters have been performed. The dependences of these characteristics on the crystal structure deformations have been investigated. The essential modification of elastic constants and bulk moduli with changes in lattice constants and stoichiometric composition has been observed. The influence of elastically stressed state of sample on X-ray diffraction intensity has been examined by using the exponential model. The model of residual stress distribution identifying in depth of wear-resistant nanostructural coating from the data of diffraction experiments has been developed.

  14. Fatigue modelling for gas nitriding

    Directory of Open Access Journals (Sweden)

    H. Weil

    2016-10-01

    Full Text Available The present study aims to develop an algorithm able to predict the fatigue lifetime of nitrided steels. Linear multi-axial fatigue criteria are used to take into account the gradients of mechanical properties provided by the nitriding process. Simulations on rotating bending fatigue specimens are made in order to test the nitrided surfaces. The fatigue model is applied to the cyclic loading of a gear from a simulation using the finite element software Ansys. Results show the positive contributions of nitriding on the fatigue strength

  15. ISOTOPIC MODEL FOR COMMERCIAL SNF BURNUP CREDIT

    Energy Technology Data Exchange (ETDEWEB)

    A.H. Wells

    2004-11-17

    The purpose of this report is to demonstrate a process for selecting bounding depletion parameters, show that they are conservative for pressurized water reactor (PWR) and boiling water reactor (BWR) spent nuclear fuel (SNF), and establish the range of burnup for which the parameters are conservative. The general range of applicability is for commercial light water reactor (LWR) SNF with initial enrichments between 2.0 and 5.0 weight percent {sup 235}U and burnups between 10 and 50 gigawatt-day per metric ton of uranium (GWd/MTU).

  16. Burn-up characteristics of ADS system utilizing the fuel composition from MOX PWRs spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Marsodi E-mail: marsodi@batan.go.id; Lasman, K.A.S.; Nishihara, K. E-mail: nishi@omega.tokai.jaeri.go.jp; Osugi, T.; Tsujimoto, K.; Marsongkohadi; Su' ud, Z. E-mail: szaki@fi.itb.ac.id

    2002-12-01

    Burn-up characteristics of accelerator-driven system, ADS has been evaluated utilizing the fuel composition from MOX PWRs spent fuel. The system consists of a high intensity proton beam accelerator, spallation target, and sub-critical reactor core. The liquid lead-bismuth, Pb-Bi, as spallation target, was put in the center of the core region. The general approach was conducted throughout the nitride fuel that allows the utilities to choose the strategy for destroying or minimizing the most dangerous high level wastes in a fast neutron spectrum. The fuel introduced surrounding the target region was the same with the composition of MOX from 33 GWd/t PWRs spent-fuel with 5 year cooling and has been compared with the fuel composition from 45 and 60 GWd/t PWRs spent-fuel with the same cooling time. The basic characteristics of the system such as burn-up reactivity swing, power density, neutron fluxes distribution, and nuclides densities were obtained from the results of the neutronics and burn-up analyses using ATRAS computer code of the Japan Atomic Energy research Institute, JAERI.

  17. Fission Gas Release in LWR Fuel Rods Exhibiting Very High Burn-Up

    DEFF Research Database (Denmark)

    Carlsen, H.

    1980-01-01

    Two UO2Zr BWR type test fuel rods were irradiated to a burn-up of about 38000 MWd/tUO2. After non-destructive characterization, the fission gas released to the internal free volume was extracted and analysed. The irradiation was simulated by means of the Danish fuel performance code WAFER-2, which...... uses an empirical gas release model combined with a strongly burn-up dependent correction term, developed by the US Nuclear Regulatory Commission. The paper presents the experimental results and the code calculations. It is concluded that the model predictions are in reasonable agreement (within 15...

  18. A PWR Thorium Pin Cell Burnup Benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, Kevan Dean; Zhao, X.; Pilat, E. E; Hejzlar, P.

    2000-05-01

    As part of work to evaluate the potential benefits of using thorium in LWR fuel, a thorium fueled benchmark comparison was made in this study between state-of-the-art codes, MOCUP (MCNP4B + ORIGEN2), and CASMO-4 for burnup calculations. The MOCUP runs were done individually at MIT and INEEL, using the same model but with some differences in techniques and cross section libraries. Eigenvalue and isotope concentrations were compared on a PWR pin cell model up to high burnup. The eigenvalue comparison as a function of burnup is good: the maximum difference is within 2% and the average absolute difference less than 1%. The isotope concentration comparisons are better than a set of MOX fuel benchmarks and comparable to a set of uranium fuel benchmarks reported in the literature. The actinide and fission product data sources used in the MOCUP burnup calculations for a typical thorium fuel are documented. Reasons for code vs code differences are analyzed and discussed.

  19. Comparison between SERPENT and MONTEBURNS codes applied to burnup calculations of a GFR-like configuration

    Energy Technology Data Exchange (ETDEWEB)

    Chersola, Davide [GeNERG – DIME/TEC, University of Genova, via all’Opera Pia 15/a, 16145 Genova (Italy); INFN, via Dodecaneso 33, 16146 Genova (Italy); Lomonaco, Guglielmo, E-mail: guglielmo.lomonaco@unige.it [GeNERG – DIME/TEC, University of Genova, via all’Opera Pia 15/a, 16145 Genova (Italy); INFN, via Dodecaneso 33, 16146 Genova (Italy); Marotta, Riccardo [GeNERG – DIME/TEC, University of Genova, via all’Opera Pia 15/a, 16145 Genova (Italy); INFN, via Dodecaneso 33, 16146 Genova (Italy); Mazzini, Guido [Centrum výzkumu Řež (Research Centre Rez), Husinec-Rez, cp. 130, 25068 Rez (Czech Republic)

    2014-07-01

    Highlights: • MC codes are widely adopted to analyze nuclear facilities, including GEN-IV reactors. • Burnup calculations are an efficient tool to test neutronic Monte Carlo codes. • In this comparison the used codes show some differences but a good agreement exists. - Abstract: This paper presents the comparison between two Monte Carlo based burnup codes: SERPENT and MONTEBURNS. Monte Carlo codes are fully and worldwide adopted to perform analyses on nuclear facilities, also in the frame of Generation IV advanced reactors simulations. Thus, faster and most powerful calculation codes are needed with the aim to analyze complex geometries and specific neutronic behaviors. Burnup calculations are an efficient tool to test neutronic Monte Carlo codes: indeed these calculations couple transport and depletion procedures, so that neutronic reactor behavior can be simulated in its totality. Comparisons have been performed on a configuration representing the Allegro MOX 75 MW{sub th} reactor proposed by the European GoFastR (Gas-cooled Fast Reactor) Project in the frame of the 7th Euratom Framework Program. Although in burnup and criticality comparisons the codes used in simulations show different calculation times and some differences in amounts and in precision (in term of statistical errors), a reasonably good agreement between them exists.

  20. Fission Product Inventory and Burnup Evaluation of the AGR-2 Irradiation by Gamma Spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Harp, Jason Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Stempien, John Dennis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Demkowicz, Paul Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO2) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. These data were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.

  1. Fission Product Inventory and Burnup Evaluation of the AGR-2 Irradiation by Gamma Spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Harp, Jason M.; Demkowicz, Paul A.; Stempien, John D.

    2016-11-01

    Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO2) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. These data were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.

  2. Burnup calculations for the HOMER-15 and SAFE-300 reactors

    Science.gov (United States)

    Amiri, Benjamin W.; Poston, David I.

    2002-01-01

    The Heatpipe Power System (HPS) is a near-term low-cost space fission power system. As the U-235 fuel of the HPS is burned, higher actinides and fission products will be produced. This will cause changes in system reactivity, radioactivity, and decay power. One potential concern is that gaseous fission products may exert excessive pressure on the fuel pin cladding. To evaluate these issues, simulations were run in MONTEBURNS. MONTEBURNS is an automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN2. This paper describes the results of these simulations, as well as how those results compare with the current experimental database of irradiated materials. .

  3. A simple gamma spectrometry method for evaluating the burnup of MTR-type HEU fuel elements

    Science.gov (United States)

    Makmal, T.; Aviv, O.; Gilad, E.

    2016-10-01

    A simple method for the evaluation of the burnup of a materials testing reactor (MTR) fuel element by gamma spectrometry is presented. The method was applied to a highly enriched uranium MTR nuclear fuel element that was irradiated in a 5 MW pool-type research reactor for a total period of 34 years. The experimental approach is based on in-situ measurements of the MTR fuel element in the reactor pool by a portable high-purity germanium detector located in a gamma cell. To corroborate the method, analytical calculations (based on the irradiation history of the fuel element) and computer simulations using a dedicated fuel cycle burnup code ORIGEN2 were performed. The burnup of the MTR fuel element was found to be 52.4±8.8%, which is in good agreement with the analytical calculations and the computer simulations. The method presented here is suitable for research reactors with either a regular or an irregular irradiation regime and for reactors with limited infrastructure and/or resources. In addition, its simplicity and the enhanced safety it confers may render this method suitable for IAEA inspectors in fuel element burnup assessments during on-site inspections.

  4. Extension and validation of the TRANSURANUS burn-up model for helium production in high burn-up LWR fuels

    Science.gov (United States)

    Botazzoli, Pietro; Luzzi, Lelio; Brémier, Stephane; Schubert, Arndt; Van Uffelen, Paul; Walker, Clive T.; Haeck, Wim; Goll, Wolfgang

    2011-12-01

    The TRANSURANUS burn-up model (TUBRNP) calculates the local concentration of the actinides, the main fission products, and 4He as a function of the radial position across a fuel rod. In this paper, the improvements in the helium production model as well as the extensions in the simulation of 238-242Pu, 241Am, 243Am and 242-245Cm isotopes are described. Experimental data used for the extended validation include new EPMA measurements of the local concentrations of Nd and Pu and recent SIMS measurements of the radial distributions of Pu, Am and Cm isotopes, both in a 3.5% enriched commercial PWR UO 2 fuel with a burn-up of 80 and 65 MWd/kgHM, respectively. Good agreement has been found between TUBRNP and the experimental data. The analysis has been complemented by detailed neutron transport calculations (VESTA code), and also revealed the need to update the branching ratio for the 241Am(n,γ) 242mAm reaction in typical PWR conditions.

  5. High burnup effects in WWER fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Smirnov, V.; Smirnov, A. [RRC Research Institute of Atomic Reactors, Dimitrovqrad (Russian Federation)

    1996-03-01

    Since 1987 at the Research Institute of Atomic Reactors, the examinations of the WWER spent fuel assemblies has been carried out. These investigations are aimed to gain information on WWER spent fuel conditions in order to validate the fuel assemblies use during the 3 and 4 year fuel cycle in the WWER-440 and WWER-1000 units. At present time, the aim is to reach an average fuel burnup of 55 MWd/kgU. According to this aim, a new investigation program on the WWER spent fuel elements is started. The main objectives of this program are to study the high burnup effects and their influence on the WWER fuel properties. This paper presented the main statistical values of the WWER-440 and WWER-1000 reactors` fuel assemblies and their fragment parameters. Average burnup of fuel in the investigated fuel assemblies was in the range of 13 to 49.7 MWd/kgU. In this case, the numer of fuel cycles was from 1 to 4 during operation of the fuel assemblies.

  6. 铁素体渗氮工艺模拟%Simulation of the Ferritic Nitriding Process

    Institute of Scientific and Technical Information of China (English)

    M.Yang; B.Yao; Y.H.Sohn; R.D.Sisson Jr

    2015-01-01

    The Lehrer diagram of pure iron is widely used to select the nitriding process of alloy steels.In the present work,the Lehrer diagram of AISI 4140 was determined for the first time by computational thermodynamics. It is demonstrated that the Lehrer diagram of AISI 4140 differs from that of pure iron significantly.Nitriding experiments for AISI 4140 were also carried out to verify the predictions from computational thermodynamics.The scanning electron microscopy results show the existence of two phases in the compound layer and the transmission electron microscopy results at the interface between the compound layer and the diffusion zone verified the coexistence of γ′-Fe4 N and ε-Fe2-3 (C,N ) phases.These experimental results agree well to the customised Lehrerdiagram of AISI 4140 constructed from computational thermodynamics.%纯铁的莱勒图被广泛用于选择合金钢的渗氮工艺。本文首次采用计算热力学来确定AISI 4140钢的莱勒图。研究表明,AISI 4140钢的莱勒图与纯铁的存在明显差异。对AISI 4140钢进行了渗氮试验,目的是验证计算热力学的预测结果。扫描电镜结果显示化合物层中存在两种相,利用透射电镜对化合物层与扩散层界面的研究结果证实了γ′-Fe4N和ε-Fe2-3(C,N)相的共存。这些试验结果与用计算热力学建立的AISI 4140钢莱勒图高度吻合。

  7. Numerical Simulation on Electrical-Thermal Properties of Gallium-Nitride-Based Light-Emitting Diodes Embedded in Board

    Directory of Open Access Journals (Sweden)

    Xing-ming Long

    2012-01-01

    Full Text Available The electrical-thermal characteristics of gallium-nitride- (GaN- based light-emitting diodes (LED, packaged by chips embedded in board (EIB technology, were investigated using a multiphysics and multiscale finite element code, COMSOL. Three-dimensional (3D finite element model for packaging structure has been developed and optimized with forward-voltage-based junction temperatures of a 9-chip EIB sample. The sensitivity analysis of the simulation model has been conducted to estimate the current and temperature distribution changes in EIB LED as the blue LED chip (substrate, indium tin oxide (ITO, packaging structure (bonding wire and chip numbers, and system condition (injection current changed. This method proved the reliability of simulated results in advance and useful material parameters. Furthermore, the method suggests that the parameter match on Shockley's equation parameters, Rs, nideal, and Is, is a potential method to reduce the current crowding effect for the EIB LED. Junction temperature decreases by approximately 3 K to 10 K can be achieved by substrate thinning, ITO, and wire bonding. The nonlinear-decreasing characteristics of total thermal resistance that decrease with an increase in chip numbers are likely to improve the thermal performance of EIB LED modules.

  8. Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, William BJ J [ORNL

    2016-01-01

    A technical basis for peak reactivity boiling water reactor (BWR) burnup credit (BUC) methods was recently generated, and the technical basis for extended BWR BUC is now being developed. In this paper, a number of effects related to extended BWR BUC are analyzed, including three major operational effects in BWRs: the coolant density axial distribution, the use of control blades during operation, and the axial burnup profile. Specifically, uniform axial moderator density profiles are analyzed and compared to previous results and an additional temporal fidelity study combing moderator density profiles for three different fuel assemblies is presented. Realistic control blade histories and cask criticality results are compared to previously generated constructed control blade histories. Finally, a preliminary study of the axial burnup profile is provided.

  9. Dynamics of 3D representation of interfaces in UV-induced chemical vapor deposition: experiments, modeling, and simulation for silicon nitride thin layers

    Science.gov (United States)

    Flicstein, Jean; Guillonneau, E.; Marquez, Jose; How Kee Chun, L. S.; Maisonneuve, D.; David, C.; Wang, Zh. Z.; Palmier, Jean F.; Courant, J. L.

    2001-06-01

    We study the surface dynamics of silicon nitride films deposited by UV-induced low pressure chemical vapor pressure. Atomic force microscopy measurements show that the surface reaches a scale invariant stationary state coherent wit the Kardar-Parisi-Zhang (KPZ) equation. Discrete geometry techniques are oriented to extra morphological characteristics of surface and bulk which corresponds to computer simulated photodeposit. This allows to determine the physical origin of KPZ scaling to be al ow value of the surface sticking probability, and connected to the surface concentration of activate charged centers, which permits to start the evaluation of the Monte Carlo-molecular dynamics simulator.

  10. Investigation and basic evaluation for ultra-high burnup fuel cladding material

    Energy Technology Data Exchange (ETDEWEB)

    Ioka, Ikuo; Nagase, Fumihisa; Futakawa, Masatoshi; Kiuchi, Kiyoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Suga, Masataka [Kokan Keisoku Co., Kawasaki, Kanagawa (Japan)

    2001-03-01

    In ultra-high burnup of the power reactor, it is an essential problem to develop the cladding with excellent durability. First, development history and approach of the safety assessment of Zircaloy for the high burnup fuel were summarized in the report. Second, the basic evaluation and investigation were carried out on the material with high practicability in order to select the candidate materials for the ultra-high burnup fuel. In addition, the basic research on modification technology of the cladding surface was carried out from the viewpoint of the addition of safety margin as a cladding. From the development history of the zirconium alloy including the Zircaloy, it is hard to estimate the results of in-pile test from those of the conventional corrosion test (out-pile test). Therefore, the development of the new testing technology that can simulate the actual environment and the elucidation of the corrosion-controlling factor of the cladding are desired. In cases of RIA (Reactivity Initiated Accident) and LOCA (Loss of Coolant Accident), it seems that the loss of ductility in zirconium alloys under heavy irradiation and boiling of high temperature water restricts the extension of fuel burnup. From preliminary evaluation on the high corrosion-resistance materials (austenitic stainless steel, iron or nickel base superalloys, titanium alloy, niobium alloy, vanadium alloy and ferritic stainless steel), stabilized austenitic stainless steels with a capability of future improvement and high-purity niobium alloys with a expectation of the good corrosion resistance were selected as candidate materials of ultra-high burnup cladding. (author)

  11. Revised SWAT. The integrated burnup calculation code system

    Energy Technology Data Exchange (ETDEWEB)

    Suyama, Kenya; Mochizuki, Hiroki [Department of Fuel Cycle Safety Research, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Kiyosumi, Takehide [The Japan Research Institute, Ltd., Tokyo (Japan)

    2000-07-01

    SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. This report shows an outline and a user's manual of revised SWAT. This revised SWAT includes expansion of functions, increasing supported machines, and correction of several bugs reported from users of previous SWAT. (author)

  12. Assessment of reactivity transient experiments with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ozer, O.; Yang, R.L.; Rashid, Y.R.; Montgomery, R.O.

    1996-03-01

    A few recent experiments aimed at determining the response of high-burnup LWR fuel during a reactivity initiated accident (RIA) have raised concerns that existing failure criteria may be inappropriate for such fuel. In particular, three experiments (SPERT CDC-859, NSRR HBO-1 and CABRI REP Na-1) appear to have resulted in fuel failures at only a fraction of the anticipated enthalpy levels. In evaluating the results of such RIA simulation experiments, however, it is necessary that the following two key considerations be taken into account: (1) Are the experiments representative of conditions that LWR fuel would experience during an in-reactor RIA event? (2) Is the fuel that is being utilized in the tests representative of the present (or anticipated) population of LWR fuel? Conducting experiments under conditions that can not occur in-reactor can trigger response modes that could not take place during in-reactor operation. Similarly, using unrepresentative fuel samples for the tests will produce failure information that is of limited relevance to commercial LWR fuel. This is particularly important for high-burnup fuel since the manner under which the test samples are base-irradiated prior to the test will impact the mechanical properties of the cladding and will therefore affect the RIA response. A good example of this effect can be seen in the results of the SPERT CDC-859 test and in the NSRR JM-4 and JM-5 tests. The conditions under which the fuel used for these tests was fabricated and/or base-irradiated prior to the RIA pulse resulted in the formation of multiple cladding defects in the form of hydride blisters. When this fuel was subjected to the RIA power pulse, it failed by developing multiple cracks that were closely correlated with the locations of the pre-existing hydride blisters. In the case of the JM tests, many of the cracks formed within the blisters themselves and did not propagate beyond the heavily hydrided regions.

  13. Analysis and finite element simulation of electromagnetic heating in the nitride MOCVD reactor

    Institute of Scientific and Technical Information of China (English)

    Li Zhi-Ming; Hao Yue; Zhang Jin-Cheng; Xu Sheng-Rui; Ni JinYu

    2009-01-01

    Electromagnetic field distribution in the vertical metal organic chemical vapour deposition (MOCVD) reactor is simulated by using the finite element method (FEM). The effects of alternating current frequency, intensity, coil turn number and the distance between the coil turns on the distribution of the Joule heat are analysed separately, and their relations to the value of Joule heat are also investigated. The temperature distribution on the suseeptor is also obtained. It is observed that the results of the simulation are in good agreement with previous measurements.

  14. Modeling and Simulation of a Gallium Nitride (GaN) Betavoltaic Energy Converter

    Science.gov (United States)

    2016-06-01

    technologies, nuclear materials detection, accelerator shielding, and dose/energy deposition in materials for medical therapies . An electron beam with...semiconductor devices have the potential to improve the efficiency of direct energy conversion and indirect energy conversion isotope batteries, making...offering higher- energy-conversion efficiency than 2-dimensional geometries. 15. SUBJECT TERMS GaN, betavoltaic, device simulation, isotope power source

  15. Assesment of advanced step models for steady state Monte Carlo burnup calculations in application to prismatic HTGR

    Directory of Open Access Journals (Sweden)

    Kępisty Grzegorz

    2015-09-01

    Full Text Available In this paper, we compare the methodology of different time-step models in the context of Monte Carlo burnup calculations for nuclear reactors. We discuss the differences between staircase step model, slope model, bridge scheme and stochastic implicit Euler method proposed in literature. We focus on the spatial stability of depletion procedure and put additional emphasis on the problem of normalization of neutron source strength. Considered methodology has been implemented in our continuous energy Monte Carlo burnup code (MCB5. The burnup simulations have been performed using the simplified high temperature gas-cooled reactor (HTGR system with and without modeling of control rod withdrawal. Useful conclusions have been formulated on the basis of results.

  16. Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian J [ORNL; Marshall, William BJ J [ORNL; Martinez-Gonzalez, Jesus S [ORNL

    2015-05-01

    Oak Ridge National Laboratory (ORNL) and the US Nuclear Regulatory Commission (NRC) have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation systems (often referred to as casks) and spent fuel pools (SFPs). This work is divided into two main phases. The first phase investigated the applicability of peak reactivity methods currently used in SFPs to transportation and storage casks and the validation of reactivity calculations and spent fuel compositions within these methods. The second phase focuses on extending BUC beyond peak reactivity. This paper documents the analysis of the effects of control blade insertion history, and moderator density and burnup axial profiles for extended BWR BUC.

  17. The Silicon / Silicon Nitride Interface and Fracture in Si: Molecular Dynamics Simulations

    Science.gov (United States)

    Bachlechner, Martina E.; Kalia, Rajiv K.; Vashishta, Priya; Ebbsjö, Ingvar

    1997-03-01

    The interface structure of a Si_3N_4(0001) film on a Si(111) substrate is studied using the molecular dynamics (MD) method. Bulk Si is described by the Stillinger-Weber potential and Si_3N4 by a combination of two-body and three-body contributions. At the interface, the charge transfer from silicon to nitrogen is taken from LCAO electronic structure calculations. Using these Si, Si_3N4 and interface interactions in MD simulations, we determine structural correlations in the interfacial regions. Results for crack propagation in silicon will also be presented.

  18. Angra 1 high burnup fuel behaviour under reactivity initiated accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Daniel de Souza; Silva, Antonio Teixeira e, E-mail: dsgomes@ipen.b, E-mail: teixeira@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    The 16x16 NGF (Next Generation Fuel) fuel assembly, comprising of highly corrosive-resistant ZIRLO clad fuel rods, been replacing the current 16x16 Standard (16STD) fuel assembly in the Angra 1, a pressurized water reactor, with a net output of 626 MWe. The 16x16 NGF fuel assemblies are designed for a peak rod average burnup of up to 75 GWd/MTU, thus improving fuel utilization and reducing spent fuel storage issues. A design basis accident, the Reactivity Initiated Accident (RIA), became a concern for a further increase in burnup as the simulated RIA tests revealed a lower enthalpy threshold for fuel failure. Two fuel performance codes, FRAPCON and FRAPTRAN, were used to predict high burnup behavior of Angra 1, during an RIA. The maximum average linear fuel rating used was 17.62 KW/m. The FRAPCON 3.4 code was applied to simulate the steady-state performance of the 16 NGF fuel rods up to a burnup of 55 GWd/MTU. With FRAPTRAN-1.4 the fuel behavior was simulated for an RIA power pulse of 4.5 ms (FHWH), and enthalpy peak of 130 Cal/g. With FRAPCON-3.4, the corrosion and hydrogen pickup characteristics of the advanced ZIRLO clad fuel rods were added to the code by modifying the actual corrosion model for Zircaloy-4 through the multiplication of empirical factors, which were appropriate to each alloy, and by means of reducing the current hydrogen pickup fraction. (author)

  19. A Simple Global View of Fuel Burnup

    Science.gov (United States)

    Sekimoto, Hiroshi

    2017-01-01

    Reactor physics and fuel burnup are discussed in order to obtain a simple global view of the effects of nuclear reactor characteristics to fuel cycle system performance. It may provide some idea of free thinking and overall vision, though it is still a small part of nuclear energy system. At the beginning of this lecture, governing equations for nuclear reactors are presented. Since the set of these equations is so big and complicated, it is simplified by imposing some extreme conditions and the nuclear equilibrium equation is derived. Some features of future nuclear equilibrium state are obtained by solving this equation. The contribution of a nucleus charged into reactor core to the system performance indexes such as criticality is worth for understanding the importance of each nuclide. It is called nuclide importance and can be evaluated by using the equations adjoint to the nuclear equilibrium equation. Examples of some importances and their application to criticalily search problem are presented.

  20. Triton burnup measurements in KSTAR using a neutron activation system

    Science.gov (United States)

    Jo, Jungmin; Cheon, MunSeong; Kim, Jun Young; Rhee, T.; Kim, Junghee; Shi, Yue-Jiang; Isobe, M.; Ogawa, K.; Chung, Kyoung-Jae; Hwang, Y. S.

    2016-11-01

    Measurements of the time-integrated triton burnup for deuterium plasma in Korea Superconducting Tokamak Advanced Research (KSTAR) have been performed following the simultaneous detection of the d-d and d-t neutrons. The d-d neutrons were measured using a 3He proportional counter, fission chamber, and activated indium sample, whereas the d-t neutrons were detected using activated silicon and copper samples. The triton burnup ratio from KSTAR discharges is found to be in the range 0.01%-0.50% depending on the plasma conditions. The measured burnup ratio is compared with the prompt loss fraction of tritons calculated with the Lorentz orbit code and the classical slowing-down time. The burnup ratio is found to increase as plasma current and classical slowing-down time increase.

  1. Calibration of burnup monitor in the Rokkasho reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Oheda, K.; Naito, H.; Hirota, M. [Japan Nuclear Fuel Ltd., Aomori (Japan); Natsume, K. [Toshiba Corp., Yokohama, Kawasaki, Kanagawa (Japan); Kumanomido, H. [Toshiba Corp., Kawasaki, Kanagawa (Japan)

    1998-07-01

    The Rokkasho Reprocessing Plant has adopted a credit for burnup in criticality control in the Spent Fuel Storage Facility (SFSF) and the Dissolution Facility. The burnup monitor system, prepared for BWR and PWR type fuel assemblies, nondestructively measures the burnup value and determines the residual U-235 enrichment in a spent fuel assembly, and criticality is controlled by the value of residual U-235 enrichment in SFSF and by the value of top 50 cm average burnup in the Dissolution Facility. The burnup monitor consists of three measurement systems; a Boss gamma-ray profile measurement system, a high resolution gamma-ray spectrometry system, and a passive neutron measurement system. The monitor sensitivity is calibrated against operator-declared burnup values through repetitive measurements of 100 spent fuel assemblies: BWR 8 X 8, PWR 14 X 14. and 17 X 17. The outline of the measurement methods, objectives of the calibration, actual calibration method, and an example of calibration performed in a demonstration experiment are presented. (author)

  2. Underestimation of nuclear fuel burnup – theory, demonstration and solution in numerical models

    Directory of Open Access Journals (Sweden)

    Gajda Paweł

    2016-01-01

    Full Text Available Monte Carlo methodology provides reference statistical solution of neutron transport criticality problems of nuclear systems. Estimated reaction rates can be applied as an input to Bateman equations that govern isotopic evolution of reactor materials. Because statistical solution of Boltzmann equation is computationally expensive, it is in practice applied to time steps of limited length. In this paper we show that simple staircase step model leads to underprediction of numerical fuel burnup (Fissions per Initial Metal Atom – FIMA. Theoretical considerations indicates that this error is inversely proportional to the length of the time step and origins from the variation of heating per source neutron. The bias can be diminished by application of predictor-corrector step model. A set of burnup simulations with various step length and coupling schemes has been performed. SERPENT code version 1.17 has been applied to the model of a typical fuel assembly from Pressurized Water Reactor. In reference case FIMA reaches 6.24% that is equivalent to about 60 GWD/tHM of industrial burnup. The discrepancies up to 1% have been observed depending on time step model and theoretical predictions are consistent with numerical results. Conclusions presented in this paper are important for research and development concerning nuclear fuel cycle also in the context of Gen4 systems.

  3. Analysis of high burnup pressurized water reactor fuel using uranium, plutonium, neodymium, and cesium isotope correlations with burnup

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jung Suk; Jeon, Young Shin; Park, Soon Dal; Ha, Yeong Keong; Song, Kyu Seok [Nuclear Chemistry Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-12-15

    The correlation of the isotopic composition of uranium, plutonium, neodymium, and cesium with the burnup for high burnup pressurized water reactor fuels irradiated in nuclear power reactors has been experimentally investigated. The total burnup was determined by Nd-148 and the fractional {sup 235}U burnup was determined by U and Pu mass spectrometric methods. The isotopic compositions of U, Pu, Nd, and Cs after their separation from the irradiated fuel samples were measured using thermal ionization mass spectrometry. The contents of these elements in the irradiated fuel were determined through an isotope dilution mass spectrometric method using {sup 233}U, {sup 242}Pu, {sup 150}Nd, and {sup 133}Cs as spikes. The activity ratios of Cs isotopes in the fuel samples were determined using gamma-ray spectrometry. The content of each element and its isotopic compositions in the irradiated fuel were expressed by their correlation with the total and fractional burnup, burnup parameters, and the isotopic compositions of different elements. The results obtained from the experimental methods were compared with those calculated using the ORIGEN-S code.

  4. An empirical formulation to describe the evolution of the high burnup structure

    Energy Technology Data Exchange (ETDEWEB)

    Lemes, Martín; Soba, Alejandro; Denis, Alicia

    2015-01-15

    In the present work the behavior of fuel pellets for LWR power reactors in the high burnup range (average burnup higher than about 45 MWd/kgU) is analyzed. For extended irradiation periods, a considerable Pu concentration is reached in the pellet periphery (rim zone), that contributes to local burnup. Gradually, a new microstructure develops in that ring, characterized by small grains and large pores as compared with those of the original material. In this region Xe is absent from the solid lattice (although it continues to be dissolved in the rest of the pellet). The porous microstructure in the pellet edge causes local changes in the mechanical and thermal properties, thus affecting the overall fuel behavior. It is generally accepted that the evolution of porosity in the high burnup structure (HBS) is determinant of the retention capacity of the fission gases rejected from the fuel matrix. This is the reason why, during the latest years a considerable effort has been devoted to characterizing the parameters that influence porosity. Although the mechanisms governing the microstructural transformation have not been completely elucidated yet, some empirical expressions can be given, and this is the intention of the present work, for representing the main physical parameters. Starting from several works published in the open literature, some mathematical expressions were developed to describe the behavior and progress of porosity at local burnup values ranging from 60 to 300 MWd/kgU. The analysis includes the interactions of different orders between pores, the growth of the pore radius by capturing vacancies, the evolution of porosity, pore number density and overpressure within the closed pores, the inventory of fission gas dissolved in the matrix and retained in the pores. The model is mathematically expressed by a system of non-linear differential equations. In the present work, results of this calculation scheme are compared with experimental data available in

  5. An empirical formulation to describe the evolution of the high burnup structure

    Science.gov (United States)

    Lemes, Martín; Soba, Alejandro; Denis, Alicia

    2015-01-01

    In the present work the behavior of fuel pellets for LWR power reactors in the high burnup range (average burnup higher than about 45 MWd/kgU) is analyzed. For extended irradiation periods, a considerable Pu concentration is reached in the pellet periphery (rim zone), that contributes to local burnup. Gradually, a new microstructure develops in that ring, characterized by small grains and large pores as compared with those of the original material. In this region Xe is absent from the solid lattice (although it continues to be dissolved in the rest of the pellet). The porous microstructure in the pellet edge causes local changes in the mechanical and thermal properties, thus affecting the overall fuel behavior. It is generally accepted that the evolution of porosity in the high burnup structure (HBS) is determinant of the retention capacity of the fission gases rejected from the fuel matrix. This is the reason why, during the latest years a considerable effort has been devoted to characterizing the parameters that influence porosity. Although the mechanisms governing the microstructural transformation have not been completely elucidated yet, some empirical expressions can be given, and this is the intention of the present work, for representing the main physical parameters. Starting from several works published in the open literature, some mathematical expressions were developed to describe the behavior and progress of porosity at local burnup values ranging from 60 to 300 MWd/kgU. The analysis includes the interactions of different orders between pores, the growth of the pore radius by capturing vacancies, the evolution of porosity, pore number density and overpressure within the closed pores, the inventory of fission gas dissolved in the matrix and retained in the pores. The model is mathematically expressed by a system of non-linear differential equations. In the present work, results of this calculation scheme are compared with experimental data available in

  6. Mechanical properties of bamboo-like boron nitride nanotubes by in situ TEM and MD simulations: strengthening effect of interlocked joint interfaces.

    Science.gov (United States)

    Tang, Dai-Ming; Ren, Cui-Lan; Wei, Xianlong; Wang, Ming-Sheng; Liu, Chang; Bando, Yoshio; Golberg, Dmitri

    2011-09-27

    Understanding the influence of interfacial structures on the nanoarchitecture mechanical properties is of particular importance for its mechanical applications. Due to a small size of constituting nanostructural units and a consequently high volume ratio of such interfacial regions, this question becomes crucial for the overall mechanical performance. Boron nitride bamboo-like nanotubes, called hereafter boron nitride nanobamboos (BNNBs), are composed of short BN nanotubular segments with specific interfaces at the bamboo-shaped joints. In this work, the mechanical properties of such structures are investigated by using direct in situ transmission electron microscopy tensile tests and molecular dynamics simulations. The mechanical properties and deformation behaviors are correlated with the interfacial structure under atomic resolution, and a geometry strengthening effect is clearly demonstrated. Due to the interlocked joint interfacial structures and compressive interfacial stresses, the deformation mechanism is switched from an interplanar sliding mode to an in-plane tensile elongation mode. As a result of such a specific geometry strengthening effect, the BNNBs show high tensile fracture strength and Young's modulus up to 8.0 and 225 GPa, respectively.

  7. Effect of burn-up and high burn-up structure on spent nuclear fuel alteration

    Energy Technology Data Exchange (ETDEWEB)

    Clarens, F.; Gonzalez-Robles, E.; Gimenez, F. J.; Casas, I.; Pablo, J. de; Serrano, D.; Wegen, D.; Glatz, J. P.; Martinez-Esparza, A.

    2009-07-01

    In this report the results of the experimental work carried out within the collaboration project between ITU-ENRESA-UPC/CTM on spent fuel (SF) covering the period 2005-2007 were presented. Studies on both RN release (Fast Release Fraction and matrix dissolution rate) and secondary phase formation were carried out by static and flow through experiments. Experiments were focussed on the study of the effect of BU with two PWR SF irradiated in commercial reactors with mean burn-ups of 48 and 60 MWd/KgU and; the effect of High Burn-up Structure (HBS) using powdered samples prepared from different radial positions. Additionally, two synthetic leaching solutions, bicarbonate and granitic bentonite ground wa ter were used. Higher releases were determined for RN from SF samples prepared from the center in comparison with the fuel from the periphery. However, within the studied range, no BU effect was observed. After one year of contact time, secondary phases were observed in batch experiments, covering the SF surface. Part of the work was performed for the Project NF-PRO of the European Commission 6th Framework Programme under contract no 2389. (Author)

  8. Crystalline boron nitride aerogels

    Science.gov (United States)

    Zettl, Alexander K.; Rousseas, Michael; Goldstein, Anna P.; Mickelson, William; Worsley, Marcus A.; Woo, Leta

    2017-04-04

    This disclosure provides methods and materials related to boron nitride aerogels. In one aspect, a material comprises an aerogel comprising boron nitride. The boron nitride has an ordered crystalline structure. The ordered crystalline structure may include atomic layers of hexagonal boron nitride lying on top of one another, with atoms contained in a first layer being superimposed on atoms contained in a second layer.

  9. Crystalline boron nitride aerogels

    Energy Technology Data Exchange (ETDEWEB)

    Zettl, Alexander K.; Rousseas, Michael; Goldstein, Anna P.; Mickelson, William; Worsley, Marcus A.; Woo, Leta

    2017-04-04

    This disclosure provides methods and materials related to boron nitride aerogels. In one aspect, a material comprises an aerogel comprising boron nitride. The boron nitride has an ordered crystalline structure. The ordered crystalline structure may include atomic layers of hexagonal boron nitride lying on top of one another, with atoms contained in a first layer being superimposed on atoms contained in a second layer.

  10. Boron nitride composites

    Energy Technology Data Exchange (ETDEWEB)

    Kuntz, Joshua D.; Ellsworth, German F.; Swenson, Fritz J.; Allen, Patrick G.

    2017-02-21

    According to one embodiment, a composite product includes: a matrix material including hexagonal boron nitride and one or more borate binders; and a plurality of cubic boron nitride particles dispersed in the matrix material. According to another embodiment, a composite product includes: a matrix material including hexagonal boron nitride and amorphous boron nitride; and a plurality of cubic boron nitride particles dispersed in the matrix material.

  11. Plasma nitriding of steels

    CERN Document Server

    Aghajani, Hossein

    2017-01-01

    This book focuses on the effect of plasma nitriding on the properties of steels. Parameters of different grades of steels are considered, such as structural and constructional steels, stainless steels and tools steels. The reader will find within the text an introduction to nitriding treatment, the basis of plasma and its roll in nitriding. The authors also address the advantages and disadvantages of plasma nitriding in comparison with other nitriding methods. .

  12. Oxidation Protection of Uranium Nitride Fuel using Liquid Phase Sintering

    Energy Technology Data Exchange (ETDEWEB)

    Dr. Paul A. Lessing

    2012-03-01

    Two methods are proposed to increase the oxidation resistance of uranium nitride (UN) nuclear fuel. These paths are: (1) Addition of USi{sub x} (e.g. U3Si2) to UN nitride powder, followed by liquid phase sintering, and (2) 'alloying' UN nitride with various compounds (followed by densification via Spark Plasma Sintering or Liquid Phase Sintering) that will greatly increase oxidation resistance. The advantages (high thermal conductivity, very high melting point, and high density) of nitride fuel have long been recognized. The sodium cooled BR-10 reactor in Russia operated for 18 years on uranium nitride fuel (UN was used as the driver fuel for two core loads). However, the potential advantages (large power up-grade, increased cycle lengths, possible high burn-ups) as a Light Water Reactor (LWR) fuel are offset by uranium nitride's extremely low oxidation resistance (UN powders oxidize in air and UN pellets decompose in hot water). Innovative research is proposed to solve this problem and thereby provide an accident tolerant LWR fuel that would resist water leaks and high temperature steam oxidation/spalling during an accident. It is proposed that we investigate two methods to increase the oxidation resistance of UN: (1) Addition of USi{sub x} (e.g. U{sub 3}Si{sub 2}) to UN nitride powder, followed by liquid phase sintering, and (2) 'alloying' UN nitride with compounds (followed by densification via Spark Plasma Sintering) that will greatly increase oxidation resistance.

  13. Methane storage in homogeneous armchair open-ended single-walled boron nitride nanotube triangular arrays: a grand canonical Monte Carlo simulation study.

    Science.gov (United States)

    Mahdizadeh, Sayyed Jalil; Tayyari, Sayyed Faramarz

    2012-06-01

    The physisorption of methane in homogeneous armchair open-ended SWBNNT triangular arrays was evaluated using grand canonical ensemble Monte Carlo simulation for tubes 11.08, 13.85, 16.62, and 19.41 Å [(8,8), (10,10), (12,12), and (14,14), respectively] in diameter, at temperatures of 273, 298, 323, and 373 K, and at fugacities of 0.5-9.0 Mpa. The intermolecular forces were modeled using the Lennard-Jones potential model. The absolute, excess, and delivery adsorption isotherms of methane were calculated for the various boron nitride nanotube arrays. The specific surface areas and the isosteric heats of adsorption, Q(st), were also studied, different isotherm models were fitted to the simulated adsorption data, and the model parameters were correlated. According to the results, it is possible to reach 108% and 140% of the US Department of Energy's target for CH(4) storage (180 v/v at 298 K and 35 bar) using the SWBNNT array with nanotubes 16.62 and 19.41 Å in diameter, respectively, as adsorbent. The results show that for a van der Waals gap of 3.4 Å, there is no interstitial adsorption except for arrays containing nanotubes with diameters of >15.8 Å. Multilayer adsorption starts to occur in arrays containing nanotubes with diameters of >16.62 Å, and the minimum pressure required for multilayer adsorption is 1.0 MPa. A brief comparison of the methane adsorption capacities of single-walled carbon and boron nitride nanotube arrays was also performed.

  14. Strategies for Application of Isotopic Uncertainties in Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I.C.

    2002-12-23

    Uncertainties in the predicted isotopic concentrations in spent nuclear fuel represent one of the largest sources of overall uncertainty in criticality calculations that use burnup credit. The methods used to propagate the uncertainties in the calculated nuclide concentrations to the uncertainty in the predicted neutron multiplication factor (k{sub eff}) of the system can have a significant effect on the uncertainty in the safety margin in criticality calculations and ultimately affect the potential capacity of spent fuel transport and storage casks employing burnup credit. Methods that can provide a more accurate and realistic estimate of the uncertainty may enable increased spent fuel cask capacity and fewer casks needing to be transported, thereby reducing regulatory burden on licensee while maintaining safety for transporting spent fuel. This report surveys several different best-estimate strategies for considering the effects of nuclide uncertainties in burnup-credit analyses. The potential benefits of these strategies are illustrated for a prototypical burnup-credit cask design. The subcritical margin estimated using best-estimate methods is discussed in comparison to the margin estimated using conventional bounding methods of uncertainty propagation. To quantify the comparison, each of the strategies for estimating uncertainty has been performed using a common database of spent fuel isotopic assay measurements for pressurized-light-water reactor fuels and predicted nuclide concentrations obtained using the current version of the SCALE code system. The experimental database applied in this study has been significantly expanded to include new high-enrichment and high-burnup spent fuel assay data recently published for a wide range of important burnup-credit actinides and fission products. Expanded rare earth fission-product measurements performed at the Khlopin Radium Institute in Russia that contain the only known publicly-available measurement for {sup 103

  15. Strategies for Application of Isotopic Uncertainties in Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I.C.

    2002-12-23

    Uncertainties in the predicted isotopic concentrations in spent nuclear fuel represent one of the largest sources of overall uncertainty in criticality calculations that use burnup credit. The methods used to propagate the uncertainties in the calculated nuclide concentrations to the uncertainty in the predicted neutron multiplication factor (k{sub eff}) of the system can have a significant effect on the uncertainty in the safety margin in criticality calculations and ultimately affect the potential capacity of spent fuel transport and storage casks employing burnup credit. Methods that can provide a more accurate and realistic estimate of the uncertainty may enable increased spent fuel cask capacity and fewer casks needing to be transported, thereby reducing regulatory burden on licensee while maintaining safety for transporting spent fuel. This report surveys several different best-estimate strategies for considering the effects of nuclide uncertainties in burnup-credit analyses. The potential benefits of these strategies are illustrated for a prototypical burnup-credit cask design. The subcritical margin estimated using best-estimate methods is discussed in comparison to the margin estimated using conventional bounding methods of uncertainty propagation. To quantify the comparison, each of the strategies for estimating uncertainty has been performed using a common database of spent fuel isotopic assay measurements for pressurized-light-water reactor fuels and predicted nuclide concentrations obtained using the current version of the SCALE code system. The experimental database applied in this study has been significantly expanded to include new high-enrichment and high-burnup spent fuel assay data recently published for a wide range of important burnup-credit actinides and fission products. Expanded rare earth fission-product measurements performed at the Khlopin Radium Institute in Russia that contain the only known publicly-available measurement for {sup 103

  16. Size Design of CdZnTe Detector Shield for Measuring Burnup of Spent Fuel

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    <正>It is important to measure the burnup of spent fuel for nuclear safeguards, burnup credit and critical safety in spent-fuel reprocessing process. The purpose of this work is designing a portable device to

  17. Power excursion analysis for BWR`s at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, D.J.; Neymoith, L.; Kohut, P. [Brookhaven National Lab., Upton, NY (United States)

    1996-03-01

    A study has been undertaken to determine the fuel enthalpy during a rod drop accident and during two thermal-hydraulic transients. The objective was to understand the consequences to high burnup fuel and the sources of uncertainty in the calculations. The analysis was done with RAMONA-4B, a computer code that models the neutron kinetics throughout the core along with the thermal-hydraulics in the core, vessel, and steamline. The results showed that the maximum fuel enthalpy in high burnup fuel will be affected by core design, initial conditions, and modeling assumptions. The important parameters in each of these categories are discussed in the paper.

  18. Study of irradiation induced restructuring of high burnup fuel - Use of computer and accelerator for fuel science and engineering -

    Energy Technology Data Exchange (ETDEWEB)

    Sataka, M.; Ishikawa, N.; Chimn, Y.; Nakamura, J.; Amaya, M. [Japan Atomic Energy Agency, Naka Gun (Japan); Iwasawa, M.; Ohnuma, T.; Sonoda, T. [Central Research Institute of Electric Power Industry, Tokyo (Japan); Kinoshita, M.; Geng, H. Y.; Chen, Y.; Kaneta, Y. [The Univ. of Tokyo, Tokyo (Japan); Yasunaga, K.; Matsumura, S.; Yasuda, K. [Kyushu Univ., Motooka (Japan); Iwase [Osaka Prefecture Univ., Osaka (Japan); Ichinomiya, T.; Nishiuran, Y. [Hokkaido Univ., Kitaku (Japan); Matzke, HJ. [Academy of Ceramics, Karlsruhe (Germany)

    2008-10-15

    In order to develop advanced fuel for future LWR reactors, trials were made to simulate the high burnup restructuring of the ceramics fuel, using accelerator irradiation out of pile and with computer simulation. The target is to reproduce the principal complex process as a whole. The reproduction of the grain subdivision (sub grain formation) was successful at experiments with sequential combined irradiation. It was made by recovery process of the accumulated dislocations, making cells and sub-boundaries at grain boundaries and pore surfaces. Details of the grain sub division mechanism is now in front of us outside of the reactor. Extensive computer science studies, first principle and molecular dynamics gave behavior of fission gas atoms and interstitial oxygen, assisting the high burnup restructuring.

  19. PWR fuel performance and burnup extension in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Yokote, M. [Kansai Electric Power Co., Inc., Osaka (Japan); Kondo, Y.; Abeta, S.

    1996-10-01

    Japanese utilities and fuel manufacturers have expanded much of their resources and efforts to maintain a reliable supply of PWR fuel for Japan. In the early 1970s, since the level of knowledge and experience of using fuel was less than now, some problems were encountered. However, their causes were investigated and countermeasures implemented, the design improved and quality control enhanced. The results can already be seen by significantly improved performance of the PWR plants now in operation, frequency of problems was quickly reduced. Since fuel reliability has been improved, the emphasis has shifted to improving economics by increasing burnup and using uranium resources effectively. The maximum discharged burnup was previously limited to 39 GWd/t and STEP1 burnup extension to 48 GWd/t has been gradually developed, while STEP2 burnup extension to 55 GWd/t is started to be demonstrated from 1996. Because resources in Japan are scarce, a policy was selected of conserving and making effective use of these resources by recycling the uranium and plutonium recovered from reactors. Consequently, significant work is being done on the development of MOX fuel and utilization of recovered uranium. (author)

  20. Need for higher fuel burnup at the Hatch Plant

    Energy Technology Data Exchange (ETDEWEB)

    Beckhman, J.T. [Georgia Power Co., Birmingham, AL (United States)

    1996-03-01

    Hatch is a BWR 4 and has been in operation for some time. The first unit became commercial about 1975. Obtaining higher burnups, or higher average discharge exposures, is nothing new at Hatch. Since we have started, the discharge exposure of the plant has increased. Now, of course, we are not approaching the numbers currently being discussed but, the average discharge exposure has increased from around 20,000 MWD/MTU in the early to mid-1980s to 34,000 MWD/MTU in 1994, I am talking about batch average values. There are also peak bundle and peak rod values. You will have to make the conversions if you think in one way or the other because I am talking in batch averages. During Hatch`s operating history we have had some problems with fuel failure. Higher burnup fuel raises a concern about how much fuel failure you are going to have. Fuel failure is, of course, an economic issue with us. Back in the early 1980s, we had a problem with crud-induced localized corrosion, known as CILC. We have gotten over that, but we had some times when it was up around 27 fuel failures a year. That is not a pleasant time to live through because it is not what you want from an economic viewpoint or any other. We have gotten that down. We have had some fuel failures recently, but they have not been related to fuel burnup or to corrosion. In fact, the number of failures has decreased from the early 1980s to the 90s even though burnup increased during that time. The fuel failures are more debris-related-type failures. In addition to increasing burnups, utilities are actively evaluating or have already incorporated power uprate and longer fuel cycles (e.g., 2-year cycles). The goal is to balance out the higher power density, longer cycles, higher burnup, and to have no leakers. Why do we as an industry want to have higher burnup fuel? That is what I want to tell you a little bit about.

  1. Dependence of control rod worth on fuel burnup

    Energy Technology Data Exchange (ETDEWEB)

    Savva, P., E-mail: savvapan@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece); Varvayanni, M., E-mail: melina@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece); Catsaros, N., E-mail: nicos@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece)

    2011-02-15

    Research highlights: Diffusion and MC calculations for rod worth dependence on burnup and Xe in reactors. One-step rod withdrawal/insertion are used for rod worth estimation. The study showed that when Xe is present the rods worth is significantly reduced. Rod worth variation with burnup depends on rod position in core. Rod worth obtained with MC code is higher than that obtained from deterministic. - Abstract: One important parameter in the design and the analysis of a nuclear reactor core is the reactivity worth of the control rods, i.e. their efficiency to absorb excess reactivity. The control rod worth is affected by parameters such as the fuel burnup in the rod vicinity, the Xe concentration in the core, the operational time of the rod and its position in the core. In the present work, two different computational approaches, a deterministic and a stochastic one, were used for the determination of the rods worth dependence on the fuel burnup level and the Xe concentration level in a conceptual, symmetric reactor core, based on the MTR fuel assemblies used in the Greek Research Reactor (GRR-1). For the deterministic approach the neutronics code system composed by the SCALE modules NITAWL and XSDRN and the diffusion code CITATION was used, while for the stochastic one the Monte Carlo code TRIPOLI was applied. The study showed that when Xe is present in the core, the rods worth is significantly reduced, while the rod worth variation with increasing burnup depends on the rods position in the core grid. The rod worth obtained with the use of the Monte Carlo code is higher than the one obtained from the deterministic code.

  2. Parametrization of the Stillinger-Weber potential for Si/N/H system and its application to simulations of silicon nitride film deposition with SiH4/NH3

    Science.gov (United States)

    Deng, Xiaodi; Song, Yixu; Li, JinChun; Pu, Yikang

    2014-02-01

    We determined the Stillinger-Weber interatomic potential parameters for Si/N/H system based on first principles density functional calculations. This new potential can be used to perform classical molecular dynamics simulation for silicon nitride deposition on Si substrate. During the first principles calculations, cluster models have been carefully and systematically chosen to make sampling of the interatomic potential supersurface more thoroughly. Global optimization method was used to fit the ab initio data into Stillinger-Weber form. We used a recursive method to perform the classical molecular dynamics simulations for silicon nitride (SiN) film growth on Si substrate with SiH4/NH3 gas mixtures. During the simulation, we could clearly observe the silicon nitride film growth progress. In this paper, we present the details of potential derivation and simulation results with different SiH4:NH3 ratios. It is demonstrated that this new potential is suitable to describe the surface reactions of the Si/N/H system and allows us to explore more complex SiN growing process such as plasma-enhanced chemical vapor deposition.

  3. Oxygen reduction reaction (ORR) on mixed oxy-nitride non-noble catalyst: Ab-initio simulation, elaboration and characterization

    Science.gov (United States)

    Seifitokaldani, Ali

    In this project, titanium oxy-nitride (TiOxN y) has been studied as a new non-noble electrocatalyst for the oxygen reduction reaction (ORR). A comprehensive comparison between four different sol-gel methods was carried out to evaluate the physicochemical and electrochemical properties of the produced electro-catalysts. Among them, a new urea-based sol-gel method (simply called U method) is introduced to prepare TiOxNy at a fairly low temperature and duration, with higher electro-catalytic activity for the ORR. The prepared electro-catalysts with different N/O ratios showed different properties from a less conductive behavior in oxygen-rich (low N/O ratio) materials to more conductive electro-catalyst behavior in nitrogen-rich (high N/O ratio) oxy-nitrides, respectively. Generally, electro-catalysts prepared by the U method had more titanium nitride in their structures than the electro-catalysts prepared by the other methods. Nevertheless, heat treatment had a key role in this phase transferring from having high oxide structure to high nitride structure. According to the elemental analysis done by energy dispersive spectroscopy (EDS), nitrogen percentage in the bulk material increased from 9 to 24 percent by increasing the temperature from 700 to 1100 °C, while the oxygen percentage was decreasing inversely. In addition, based on the X-ray diffraction (XRD) data, in the case of U method, the TiN characteristic peaks were obvious, even at lower temperatures. Increasing the temperature also made the peaks much sharper indicating the growth of the crystallite size. The calculated crystallite size showed the crystallite size of samples prepared by U method (20 to 40 nm) was almost in the same range of the TiN crystallite size, but the crystallite size of the samples prepared by the other sol-gel methods (40 to 60 nm) was in the same range of the TiO2 crystallite size. Scanning electron microscopy (SEM) and B.E.T. surface area analyzer were used to evaluate the particle

  4. THE INVESTIGATION OF BURNUP CHARACTERISTICS USING THE SERPENT MONTE CARLO CODE FOR A SODIUM COOLED FAST REACTOR

    Directory of Open Access Journals (Sweden)

    MEHMET E. KORKMAZ

    2014-06-01

    Full Text Available In this research, we investigated the burnup characteristics and the conversion of fertile 232Th into fissile 233U in the core of a Sodium-Cooled Fast Reactor (SFR. The SFR fuel assemblies were designed for burning 232Th fuel (fuel pin 1 and 233U fuel (fuel pin 2 and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method and TTA (Transmutation Trajectory Analysis method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff was between 0.964 and 0.954 and peaking factor is 1.88867.

  5. The investigation of burnup characteristics using the serpent Monte Carlo code for a sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Korkmaz, Mehmet E.; Agar, Osman [Karamanoglu Mehmetbey University, Faculty of Kamil Oezdag Science, Karaman (Turkmenistan)

    2014-06-15

    In this research, we investigated the burnup characteristics and the conversion of fertile {sup 232}Th into fissile {sup 233}U in the core of a Sodium-Cooled Fast Reactor (SFR). The SFR fuel assemblies were designed for burning {sup 232}Th fuel (fuel pin 1) and {sup 233}U fuel (fuel pin 2) and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method) and TTA (Transmutation Trajectory Analysis) method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff) was between 0.964 and 0.954 and peaking factor is 1.88867.

  6. A multi-platform linking code for fuel burnup and radiotoxicity analysis

    Science.gov (United States)

    Cunha, R.; Pereira, C.; Veloso, M. A. F.; Cardoso, F.; Costa, A. L.

    2014-02-01

    A linking code between ORIGEN2.1 and MCNP has been developed at the Departamento de Engenharia Nuclear/UFMG to calculate coupled neutronic/isotopic results for nuclear systems and to produce a large number of criticality, burnup and radiotoxicity results. In its previous version, it evaluated the isotopic composition evolution in a Heat Pipe Power System model as well as the radiotoxicity and radioactivity during lifetime cycles. In the new version, the code presents features such as multi-platform execution and automatic results analysis. Improvements made in the code allow it to perform simulations in a simpler and faster way without compromising accuracy. Initially, the code generates a new input for MCNP based on the decisions of the user. After that, MCNP is run and data, such as recoverable energy per prompt fission neutron, reaction rates and keff, are automatically extracted from the output and used to calculate neutron flux and cross sections. These data are then used to construct new ORIGEN inputs, one for each cell in the core. Each new input is run on ORIGEN and generates outputs that represent the complete isotopic composition of the core on that time step. The results show good agreement between GB (Coupled Neutronic/Isotopic code) and Monteburns (Automated, Multi-Step Monte Carlo Burnup Code System), developed by the Los Alamos National Laboratory.

  7. ATR PDQ and MCWO Fuel Burnup Analysis Codes Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    G.S. Chang; P. A. Roth; M. A. Lillo

    2009-11-01

    The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is being studied to determine the feasibility of converting it from the highly enriched Uranium (HEU) fuel that is currently uses to low enriched Uranium (LEU) fuel. In order to achieve this goal, it would be best to qualify some different computational methods than those that have been used at ATR for the past 40 years. This paper discusses two methods of calculating the burnup of ATR fuel elements. The existing method, that uses the PDQ code, is compared to a modern method that uses A General Monte Carlo N-Particle Transport Code (MCNP) combined with the Origen2.2 code. This modern method, MCNP with ORIGEN2.2 (MCWO), is found to give excellent agreement with the existing method (PDQ). Both of MCWO and PDQ are also in a very good agreement to the 235U burnup data generated by an analytical method.

  8. Propagation of Uncertainty in System Parameters of a LWR Model by Sampling MCNPX Calculations - Burnup Analysis

    Science.gov (United States)

    Campolina, Daniel de A. M.; Lima, Claubia P. B.; Veloso, Maria Auxiliadora F.

    2014-06-01

    For all the physical components that comprise a nuclear system there is an uncertainty. Assessing the impact of uncertainties in the simulation of fissionable material systems is essential for a best estimate calculation that has been replacing the conservative model calculations as the computational power increases. The propagation of uncertainty in a simulation using a Monte Carlo code by sampling the input parameters is recent because of the huge computational effort required. In this work a sample space of MCNPX calculations was used to propagate the uncertainty. The sample size was optimized using the Wilks formula for a 95th percentile and a two-sided statistical tolerance interval of 95%. Uncertainties in input parameters of the reactor considered included geometry dimensions and densities. It was showed the capacity of the sampling-based method for burnup when the calculations sample size is optimized and many parameter uncertainties are investigated together, in the same input.

  9. Impact of Reactor Operating Parameters on Cask Reactivity in BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Betzler, Benjamin R [ORNL; Ade, Brian J [ORNL

    2017-01-01

    This paper discusses the effect of reactor operating parameters used in fuel depletion calculations on spent fuel cask reactivity, with relevance for boiling-water reactor (BWR) burnup credit (BUC) applications. Assessments that used generic BWR fuel assembly and spent fuel cask configurations are presented. The considered operating parameters, which were independently varied in the depletion simulations for the assembly, included fuel temperature, bypass water density, specific power, and operating history. Different operating history scenarios were considered for the assembly depletion to determine the effect of relative power distribution during the irradiation cycles, as well as the downtime between cycles. Depletion, decay, and criticality simulations were performed using computer codes and associated nuclear data within the SCALE code system. Results quantifying the dependence of cask reactivity on the assembly depletion parameters are presented herein.

  10. Impacts of burnup-dependent swelling of metallic fuel on the performance of a compact breed-and-burn fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hartanto, Donny; Heo, Woong; Kim, Chi Hyung; Kim, Yong Hee [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology (KAIST), Daejeon (Korea, Republic of)

    2016-04-15

    The U-Zr or U-TRU-Zr cylindrical metallic fuel slug used in fast reactors is known to swell significantly and to grow during irradiation. In neutronics simulations of metallic-fueled fast reactors, it is assumed that the slug has swollen and contacted cladding, and the bonding sodium has been removed from the fuel region. In this research, a realistic burnup-dependent fuel-swelling simulation was performed using Monte Carlo code McCARD for a single-batch compact sodium-cooled breed-and-burn reactor by considering the fuel-swelling behavior reported from the irradiation test results in EBR-II. The impacts of the realistic burnup-dependent fuel swelling are identified in terms of the reactor neutronics performance, such as core lifetime, conversion ratio, axial power distribution, and local burnup distributions. It was found that axial fuel growth significantly deteriorated the neutron economy of a breed-and-burn reactor and consequently impaired its neutronics performance. The bonding sodium also impaired neutron economy, because it stayed longer in the blanket region until the fuel slug reached 2% burnup.

  11. High resolution transmission electron microscopy and molecular simulation analysis of ε-Fe{sub 2}-{sub 3}N and γ-Fe{sub 4}N formation for a nitrided 4140 steel

    Energy Technology Data Exchange (ETDEWEB)

    Bejar, L.; Medina-Flores, A.; Alfonso, I.; Carreon, H.; Solis, J.

    2015-07-01

    High Resolution Transmission Electron Microscopy (HRTEM) and molecular simulation techniques were applied to analyze phase formation in the post-discharge microwave nitriding process of a 4140 steel. The results showed the formation of ε-Fe{sub 2}-{sub 3}N and γ-Fe{sub 4}N. The morphology of theγ-Fe{sub 4}N phase was composed by cubic nanoparticles that grows in the <0 0 1> axis direction. Otherwise, the ε-Fe{sub 2}-{sub 3}N crystal is associated to hexagonal faces in the [0 0 1] direction which grows forming angles of 120 degree centigrade. The comparison between experimental and analytical results allowed a better understanding of the phases formed during this nitriding process. (Author)

  12. Models for fuel rod behaviour at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Jernkvist, Lars O.; Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park, Uppsala (Sweden)

    2004-12-01

    This report deals with release of fission product gases and irradiation-induced restructuring in uranium dioxide nuclear fuel. Waterside corrosion of zirconium alloy clad tubes to light water reactor fuel rods is also discussed. Computational models, suitable for implementation in the FRAPCON-3.2 computer code, are proposed for these potentially life-limiting phenomena. Hence, an integrated model for the calculation or thermal fission gas release by intragranular diffusion, gas trapping in grain boundaries, irradiation-induced re-solution, grain boundary saturation, and grain boundary sweeping in UO{sub 2} fuel, under time varying temperature loads, is formulated. After a brief review of the status of thermal fission gas release modelling, we delineate the governing equations for the aforementioned processes. Grain growth kinetic modelling is briefly reviewed and pertinent data on grain growth of high burnup fuel obtained during power ramps in the Third Risoe Fission Gas Release Project are evaluated. Sample computations are performed, which clearly show the connection between fission gas release and gram growth as a function of time at different isotherms. Models are also proposed for the restructuring of uranium dioxide fuel at high burnup, the so-called rim formation, and its effect on fuel porosity build-up, fuel thermal conductivity and fission gas release. These models are assessed by use of recent experimental data from the High Burnup Rim Project, as well as from post irradiation examinations of high-burnup fuel, irradiated in power reactors. Moreover, models for clad oxide growth and hydrogen pickup in PWRs, applicable to Zircaloy-4, ZIRLO or M5 cladding, are formulated, based on recent in-reactor corrosion data for high-burnup fuel rods. Our evaluation of these data indicates that the oxidation rate of ZIRLO-type materials is about 20% lower than for standard Zircaloy-4 cladding under typical PWR conditions. Likewise, the oxidation rate of M5 seems to be

  13. ZeroFlow - new, environmentally friendly method of controlled gas nitriding used for selected car parts

    Science.gov (United States)

    Kowalska, J.; Małdziński, L.

    2016-09-01

    This article presents new method of controlled gas nitriding called ZeroFlow, which is used for nitriding of selected car parts, such as crankshafts, camshafts, piston rings, poppet valve springs and discs, piston pins or nozzles for unit injectors. This article will discuss the essence of controlled gas nitriding process, with an emphasis on the influence of process parameters on results of nitriding process. This information are the basis to understand the issue of the kinetics of nitrided layer growth, and as it follows - for its practical application in designing, regulation and control of nitriding processes using simulation models (simulator of the kinetics of nitrided layer growth). This article will also present the simulator of the kinetics of nitrided layer growth, which supports nitriding using ZeroFlow method - through the use of simulator layers are obtained in the shortest possible time, which is connected with the lowest energy consumption; therefore, nitriding process using ZeroFlow method and simulator of the kinetics of nitrided layer growth is both economical and environmentally friendly.

  14. Local heating with titanium nitride nanoparticles

    DEFF Research Database (Denmark)

    Guler, Urcan; Ndukaife, Justus C.; Naik, Gururaj V.;

    2013-01-01

    We investigate the feasibility of titanium nitride (TiN) nanoparticles as local heat sources in the near infrared region, focusing on biological window. Experiments and simulations provide promising results for TiN, which is known to be bio-compatible.......We investigate the feasibility of titanium nitride (TiN) nanoparticles as local heat sources in the near infrared region, focusing on biological window. Experiments and simulations provide promising results for TiN, which is known to be bio-compatible....

  15. Atomistic comparative study of VUV photodeposited silicon nitride on InP(100) by simulation and atomic force microscopy

    Science.gov (United States)

    Flicstein, J.; Guillonneau, E.; Marquez, J.; How Kee Chun, L. S.; Maisonneuve, D.; David, C.; Wang, Zh.; Palmier, J. F.; Courant, J. L.

    2000-02-01

    We report on an accurate validation of a new Monte Carlo three-dimensional model. Simulations up to 1200 Å layer thickness have been carried out for amorphous thin film layers of SiN:H deposited at low temperature (400-650 K) on (100) InP, by vacuum ultraviolet (VUV, ˜185 nm)-induced chemical vapor deposition (CVD). The computer simulations in the mesoscopic-submicronic range are compared with atomic force microscopy and index of refraction measurements. The reconstituted surface roughness and the voids discrete representations of the bulk are found to be in good agreement with these measurements. Simultaneously at around 450 K (at ˜175°C), thermal characteristic evolution of the both surface roughness and bulk porosity showed a transition from rough to smooth deposition and from low to high density.

  16. Modeling and simulation of the deposition/relaxation processes of polycrystalline diatomic structures of metallic nitride films

    Science.gov (United States)

    García, M. F.; Restrepo-Parra, E.; Riaño-Rojas, J. C.

    2015-05-01

    This work develops a model that mimics the growth of diatomic, polycrystalline thin films by artificially splitting the growth into deposition and relaxation processes including two stages: (1) a grain-based stochastic method (grains orientation randomly chosen) is considered and by means of the Kinetic Monte Carlo method employing a non-standard version, known as Constant Time Stepping, the deposition is simulated. The adsorption of adatoms is accepted or rejected depending on the neighborhood conditions; furthermore, the desorption process is not included in the simulation and (2) the Monte Carlo method combined with the metropolis algorithm is used to simulate the diffusion. The model was developed by accounting for parameters that determine the morphology of the film, such as the growth temperature, the interacting atomic species, the binding energy and the material crystal structure. The modeled samples exhibited an FCC structure with grain formation with orientations in the family planes of , and . The grain size and film roughness were analyzed. By construction, the grain size decreased, and the roughness increased, as the growth temperature increased. Although, during the growth process of real materials, the deposition and relaxation occurs simultaneously, this method may perhaps be valid to build realistic polycrystalline samples.

  17. A feasibility and optimization study to determine cooling time and burnup of advanced test reactor fuels using a nondestructive technique

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, Jorge [Univ. of Utah, Salt Lake City, UT (United States)

    2013-12-01

    for ATR applications the technique was tested using one-isotope, multi-isotope and fuel simulated sources. Burnup calibrations were perfomed using convoluted and deconvoluted data. The calibrations results showed burnup prediction by this method improves using deconvolution. The final stage of the deconvolution method development was to perform an irradiation experiment in order to create a surrogate fuel source to test the deconvolution method using experimental data. A conceptual design of the fuel scan system is path forward using the rugged LaBr3 detector in an above the water configuration and deconvolution algorithms.

  18. Methods of forming boron nitride

    Science.gov (United States)

    Trowbridge, Tammy L; Wertsching, Alan K; Pinhero, Patrick J; Crandall, David L

    2015-03-03

    A method of forming a boron nitride. The method comprises contacting a metal article with a monomeric boron-nitrogen compound and converting the monomeric boron-nitrogen compound to a boron nitride. The boron nitride is formed on the same or a different metal article. The monomeric boron-nitrogen compound is borazine, cycloborazane, trimethylcycloborazane, polyborazylene, B-vinylborazine, poly(B-vinylborazine), or combinations thereof. The monomeric boron-nitrogen compound is polymerized to form the boron nitride by exposure to a temperature greater than approximately 100.degree. C. The boron nitride is amorphous boron nitride, hexagonal boron nitride, rhombohedral boron nitride, turbostratic boron nitride, wurzite boron nitride, combinations thereof, or boron nitride and carbon. A method of conditioning a ballistic weapon and a metal article coated with the monomeric boron-nitrogen compound are also disclosed.

  19. Methods of forming boron nitride

    Energy Technology Data Exchange (ETDEWEB)

    Trowbridge, Tammy L; Wertsching, Alan K; Pinhero, Patrick J; Crandall, David L

    2015-03-03

    A method of forming a boron nitride. The method comprises contacting a metal article with a monomeric boron-nitrogen compound and converting the monomeric boron-nitrogen compound to a boron nitride. The boron nitride is formed on the same or a different metal article. The monomeric boron-nitrogen compound is borazine, cycloborazane, trimethylcycloborazane, polyborazylene, B-vinylborazine, poly(B-vinylborazine), or combinations thereof. The monomeric boron-nitrogen compound is polymerized to form the boron nitride by exposure to a temperature greater than approximately 100.degree. C. The boron nitride is amorphous boron nitride, hexagonal boron nitride, rhombohedral boron nitride, turbostratic boron nitride, wurzite boron nitride, combinations thereof, or boron nitride and carbon. A method of conditioning a ballistic weapon and a metal article coated with the monomeric boron-nitrogen compound are also disclosed.

  20. MONTE-CARLO BURNUP CALCULATION UNCERTAINTY QUANTIFICATION AND PROPAGATION DETERMINATION

    Energy Technology Data Exchange (ETDEWEB)

    Nichols, T.; Sternat, M.; Charlton, W.

    2011-05-08

    MONTEBURNS is a Monte-Carlo depletion routine utilizing MCNP and ORIGEN 2.2. Uncertainties exist in the MCNP transport calculation, but this information is not passed to the depletion calculation in ORIGEN or saved. To quantify this transport uncertainty and determine how it propagates between burnup steps, a statistical analysis of a multiple repeated depletion runs is performed. The reactor model chosen is the Oak Ridge Research Reactor (ORR) in a single assembly, infinite lattice configuration. This model was burned for a 25.5 day cycle broken down into three steps. The output isotopics as well as effective multiplication factor (k-effective) were tabulated and histograms were created at each burnup step using the Scott Method to determine the bin width. It was expected that the gram quantities and k-effective histograms would produce normally distributed results since they were produced from a Monte-Carlo routine, but some of results do not. The standard deviation at each burnup step was consistent between fission product isotopes as expected, while the uranium isotopes created some unique results. The variation in the quantity of uranium was small enough that, from the reaction rate MCNP tally, round off error occurred producing a set of repeated results with slight variation. Statistical analyses were performed using the {chi}{sup 2} test against a normal distribution for several isotopes and the k-effective results. While the isotopes failed to reject the null hypothesis of being normally distributed, the {chi}{sup 2} statistic grew through the steps in the k-effective test. The null hypothesis was rejected in the later steps. These results suggest, for a high accuracy solution, MCNP cell material quantities less than 100 grams and greater kcode parameters are needed to minimize uncertainty propagation and minimize round off effects.

  1. Imitators of plutonium and americium in a mixed uranium- plutonium nitride fuel

    Science.gov (United States)

    Nikitin, S. N.; Shornikov, D. P.; Tarasov, B. A.; Baranov, V. G.; Burlakova, M. A.

    2016-04-01

    Uranium nitride and mix uranium nitride (U-Pu)N is most popular nuclear fuel for Russian Fast Breeder Reactor. The works in hot cells associated with the radiation exposure of personnel and methodological difficulties. To know the main physical-chemical properties of uranium-plutonium nitride it necessary research to hot cells. In this paper, based on an assessment of physicochemical and thermodynamic properties of selected simulators Pu and Am. Analogues of Pu is are Ce and Y, and analogues Am - Dy. The technique of obtaining a model nitride fuel based on lanthanides nitrides and UN. Hydrogenation-dehydrogenation- nitration method of derived powders nitrides uranium, cerium, yttrium and dysprosium, held their mixing, pressing and sintering, the samples obtained model nitride fuel with plutonium and americium imitation. According to the results of structural studies have shown that all the samples are solid solution nitrides rare earth (REE) elements in UN.

  2. New results from the NSRR experiments with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Fuketa, Toyoshi; Ishijima, Kiyomi; Mori, Yukihide [Japan Atomic Research Institute, Toaki, Ibaraki (Japan)] [and others

    1996-03-01

    Results obtained in the NSRR power burst experiments with irradiated PWR fuel rods with fuel burnup up to 50 MWd/kgU are described and discussed in this paper. Data concerning test method, test fuel rod, pulse irradiation, transient records during the pulse and post irradiation examination are described, and interpretations and discussions on fission gas release and fuel pellet fragmentation are presented. During the pulse-irradiation experiment with 50 MWd/kgU PWR fuel rod, the fuel rod failed at considerably low energy deposition level, and large amount of fission gas release and fragmentation of fuel pellets were observed.

  3. Determination of IRT-2M fuel burnup by gamma spectrometry.

    Science.gov (United States)

    Koleška, Michal; Viererbl, Ladislav; Marek, Milan; Ernest, Jaroslav; Šunka, Michal; Vinš, Miroslav

    2016-01-01

    A spectrometric system was developed for evaluating spent fuel in the LVR-15 research reactor, which employs highly enriched (36%) IRT-2M-type fuel. Such system allows the measurement of detailed fission product profiles. Within these measurements, nuclides such as (137)Cs, (134)Cs, (144)Ce, (106)Ru and (154)Eu may be detected in fuel assemblies with different cooling times varying between 1.67 and 7.53 years. Burnup calculations using the MCNPX Monte Carlo code data showed good agreement with measurements, though some discrepancies were observed in certain regions. These discrepancies are attributed to the evaluation of irradiation history, reactor regulation pattern and buildup schemes.

  4. Thermodynamic analysis for high burn-up fuel internal chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Fuji, Kensho; Kyoh, Bunkei [Kinki Univ., Higashi-Osaka, Osaka (Japan). Faculty of Science and Technology

    1997-09-01

    Chemical states of fission products and actinide elements in high burn-up LWR fuel pellets have been analyzed thermodynamically using the computer program SOLGASMIX-PV. Calculations with this computer code have been performed for a complex multi-component system, which comprises 54 chemical species. The analysis shows that neither alkali nor alkaline-earth uranates are formed, but alkali and alkaline-earth molybdates exist in irradiated LWR fuel pellets in contrast with their post irradiation examinations. These molybdates tend to increase with increasing oxygen potential in the fuel under operating conditions, whereas the zirconates decrease. (author)

  5. Stability characteristics and structural properties of single- and double-walled boron-nitride nanotubes under physical adsorption of Flavin mononucleotide (FMN) in aqueous environment using molecular dynamics simulations

    Energy Technology Data Exchange (ETDEWEB)

    Ansari, R., E-mail: r_ansari@guilan.ac.ir; Ajori, S., E-mail: Shahram_ajori1366@yahoo.com; Ameri, A.

    2016-03-15

    Graphical abstract: Structural properties and stability characteristics of single- and double-walled boron-nitride nanotubes functionalized with Flavin mononucleotide (FMN) in aqueous environment are investigated employing molecular dynamics simulations. - Highlights: • Structural and buckling analysis of boron-nitride nanotubes under physical adsorption of Flavin mononucleotide (FMN). • Gyration radius increases linearly as the weight percentage of FMN increases. • Presence of water molecules results in more expansion of FMN around BNNTs. • Critical buckling force of functionalized BNNTs is higher than that of pure BNNTs. • The critical strain of functionalized BNNTs is found to be lower than that of pure ones. - Abstract: The non-cytotoxic properties of Boron-nitride nanotubes (BNNTs) and the ability of stable interaction with biomolecules make them so promising for biological applications. In this research, molecular dynamics (MD) simulations are performed to investigate the structural properties and stability characteristics of single- and double-walled BNNTs under physical adsorption of Flavin mononucleotide (FMN) in vacuum and aqueous environments. According to the simulation results, gyration radius increases by rising the weight percentage of FMN. Also, the results demonstrate that critical buckling force of functionalized BNNTs increases in vacuum. Moreover, it is observed that by increasing the weight percentage of FMN, critical force of functionalized BNNTs rises. By contrast, critical strain reduces by functionalization of BNNTs in vacuum. Considering the aqueous environment, it is observed that gyration radius and critical buckling force of functionalized BNNTs increase more considerably than those of functionalized BNNTs in vacuum, whereas the critical strains approximately remain unchanged.

  6. Silicon nitride equation of state

    Science.gov (United States)

    Brown, Robert C.; Swaminathan, Pazhayannur K.

    2017-01-01

    This report presents the development of a global, multi-phase equation of state (EOS) for the ceramic silicon nitride (Si3N4).1 Structural forms include amorphous silicon nitride normally used as a thin film and three crystalline polymorphs. Crystalline phases include hexagonal α-Si3N4, hexagonal β-Si3N4, and the cubic spinel c-Si3N4. Decomposition at about 1900 °C results in a liquid silicon phase and gas phase products such as molecular nitrogen, atomic nitrogen, and atomic silicon. The silicon nitride EOS was developed using EOSPro which is a new and extended version of the PANDA II code. Both codes are valuable tools and have been used successfully for a variety of material classes. Both PANDA II and EOSPro can generate a tabular EOS that can be used in conjunction with hydrocodes. The paper describes the development efforts for the component solid phases and presents results obtained using the EOSPro phase transition model to investigate the solid-solid phase transitions in relation to the available shock data that have indicated a complex and slow time dependent phase change to the c-Si3N4 phase. Furthermore, the EOSPro mixture model is used to develop a model for the decomposition products; however, the need for a kinetic approach is suggested to combine with the single component solid models to simulate and further investigate the global phase coexistences.

  7. Fabrication characteristics of dry process fuel with a variation of fuel burn-ups

    Energy Technology Data Exchange (ETDEWEB)

    Park, Geun Il; Kim, W. K.; Lee, J. W. [and others

    2004-11-01

    Fabrication characteristics of the dry processed fuel with a variation of fuel burn-ups in a range of 27,300 to 65,000 MWD/tU were experimentally evaluated. Density comparison of powders which were fabricated from oxidation, OREOX and milling processes at same process conditions was performed with a function of fuel burn-ups respectively. The influence of fuel burn-ups on sintering characteristics of dry processed fuel was studied by comparing the density change of sintered pellet as well as green pellet. Weight gain by fuel oxidation to U{sub 3}O{sub 8} showed semi-linear dependence with increasing fuel burn-ups. OREOX powder density increased up to 3.7 g/cm{sup 3} at high burn-up fuel, and the density of milled powder with fuel burn-ups represented almost similar value of 3.2{+-}0.2 g/cm{sup 3}. Also, the green pellet density compacted by 120 MPa decreased smoothly with increasing fuel burn-ups, and the density change of sintered pellet showed the similar trend as green pellet. The sintered density of pellet in a range of 27,000 to 40,000 MWD/tU was found to be more 95% of Theoretical Density(T.D.), but the sintered pellet density fabricated from high burn-up fuel showed a range of 92 % to 93% of T.D.

  8. Evaluation of Isotopic Measurements and Burn-up Value of Sample GU3 of ARIANE Project

    Energy Technology Data Exchange (ETDEWEB)

    Tore, C.; Rodriguez Rivada, A.

    2014-07-01

    Estimation of the burn-up value of irradiated fuel and its isotopic composition are important for criticality analysis, spent fuel management and source term estimation. The practical way to estimate the irradiated fuel composition and burn.up value is calculation with validated code and nuclear data. Such validation of the neutronic codes and nuclear data requires the benchmarking with measured values. (Author)

  9. MTR core loading pattern optimization using burnup dependent group constants

    Directory of Open Access Journals (Sweden)

    Iqbal Masood

    2008-01-01

    Full Text Available A diffusion theory based MTR fuel management methodology has been developed for finding superior core loading patterns at any stage for MTR systems, keeping track of burnup of individual fuel assemblies throughout their history. It is based on using burnup dependent group constants obtained by the WIMS-D/4 computer code for standard fuel elements and control fuel elements. This methodology has been implemented in a computer program named BFMTR, which carries out detailed five group diffusion theory calculations using the CITATION code as a subroutine. The core-wide spatial flux and power profiles thus obtained are used for calculating the peak-to-average power and flux-ratios along with the available excess reactivity of the system. The fuel manager can use the BFMTR code for loading pattern optimization for maximizing the excess reactivity, keeping the peak-to-average power as well as flux-ratio within constraints. The results obtained by the BFMTR code have been found to be in good agreement with the corresponding experimental values for the equilibrium core of the Pakistan Research Reactor-1.

  10. Neutron transport-burnup code MCORGS and its application in fusion fission hybrid blanket conceptual research

    Science.gov (United States)

    Shi, Xue-Ming; Peng, Xian-Jue

    2016-09-01

    Fusion science and technology has made progress in the last decades. However, commercialization of fusion reactors still faces challenges relating to higher fusion energy gain, irradiation-resistant material, and tritium self-sufficiency. Fusion Fission Hybrid Reactors (FFHR) can be introduced to accelerate the early application of fusion energy. Traditionally, FFHRs have been classified as either breeders or transmuters. Both need partition of plutonium from spent fuel, which will pose nuclear proliferation risks. A conceptual design of a Fusion Fission Hybrid Reactor for Energy (FFHR-E), which can make full use of natural uranium with lower nuclear proliferation risk, is presented. The fusion core parameters are similar to those of the International Thermonuclear Experimental Reactor. An alloy of natural uranium and zirconium is adopted in the fission blanket, which is cooled by light water. In order to model blanket burnup problems, a linkage code MCORGS, which couples MCNP4B and ORIGEN-S, is developed and validated through several typical benchmarks. The average blanket energy Multiplication and Tritium Breeding Ratio can be maintained at 10 and 1.15 respectively over tens of years of continuous irradiation. If simple reprocessing without separation of plutonium from uranium is adopted every few years, FFHR-E can achieve better neutronic performance. MCORGS has also been used to analyze the ultra-deep burnup model of Laser Inertial Confinement Fusion Fission Energy (LIFE) from LLNL, and a new blanket design that uses Pb instead of Be as the neutron multiplier is proposed. In addition, MCORGS has been used to simulate the fluid transmuter model of the In-Zinerater from Sandia. A brief comparison of LIFE, In-Zinerater, and FFHR-E will be given.

  11. Triton burnup study using scintillating fiber detector on JT-60U

    Energy Technology Data Exchange (ETDEWEB)

    Harano, Hideki [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1997-09-01

    The DT fusion reactor cannot be realized without knowing how the fusion-produced 3.5 MeV {alpha} particles behave. The {alpha} particles` behavior can be simulated using the 1 MeV triton. To investigate the 1 MeV triton`s behavior, a new type of directional 14 MeV neutron detector, scintillating fiber (Sci-Fi) detector has been developed and installed on JT-60U in the cooperation with LANL as part of a US-Japan collaboration. The most remarkable feature of the Sci-Fi detector is that the plastic scintillating fibers are employed for the neutron sensor head. The Sci-Fi detector measures and extracts the DT neutrons from the fusion radiation field in high time resolution (10 ms) and wide dynamic range (3 decades). Triton burnup analysis code TBURN has been made in order to analyze the time evolution of DT neutron emission rate obtained by the Sci-Fi detector. The TBURN calculations reproduced the measurements fairly well, and the validity of the calculation model that the slowing down of the 1 MeV triton was classical was confirmed. The Sci-Fi detector`s directionality indicated the tendency that the DT neutron emission profile became more and more peaked with the time progress. In this study, in order to examine the effect of the toroidal field ripple on the triton burnup, R{sub p}-scan and n{sub e}-scan experiments have been performed. The R{sub p}-scan experiment indicates that the triton`s transport was increased as the ripple amplitude over the triton became larger. In the n{sub e}-scan experiment, the DT neutron emission showed the characteristic changes after the gas puffing injection. It was theoretically confirmed that the gas puffing was effective for the collisionality scan. (J.P.N.) 127 refs.

  12. Calibration of burnup monitor of spent nuclear fuel installed at Rokkasho reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Oeda, Kaoru; Matoba, Masaru; Wakabayashi, Genichiro [Kyushu Univ., Fukuoka (Japan). Faculty of Engineering; Naito, Hirofumi; Hirota, Masanari [Nuclear Fuel Industries Ltd., Tokyo (Japan); Morizaki, Hidetoshi; Kumanomido, Hironori; Natsume, Koichiro [Toshiba Corp., Tokyo (Japan)

    2001-05-01

    The spent nuclear fuel storage pool of Rokkasho reprocessing plant adopts the burnup credit' conception. Spent fuel assemblies are measured every one by one, by burnup monitors, and stored to a storage rack which is designed with specified residual enrichment. For nuclear criticality control, it is necessary for the burnup monitor that the measured value includes a kind of margin, which consists of errors of the monitor. In this paper, we describe the error of the burnup monitors, and the way of taking of the margin. From the result of calibration of the burnup monitor carried out from July through November, 1999, we describe that the way of taking of the margin is validated. And comments about possibility of error reduction are remarked. (author)

  13. Detailed description and user`s manual of high burnup fuel analysis code EXBURN-I

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Saitou, Hiroaki

    1997-11-01

    EXBURN-I has been developed for the analysis of LWR high burnup fuel behavior in normal operation and power transient conditions. In the high burnup region, phenomena occur which are different in quality from those expected for the extension of behaviors in the mid-burnup region. To analyze these phenomena, EXBURN-I has been formed by the incorporation of such new models as pellet thermal conductivity change, burnup-dependent FP gas release rate, and cladding oxide layer growth to the basic structure of low- and mid-burnup fuel analysis code FEMAXI-IV. The present report describes in detail the whole structure of the code, models, and materials properties. Also, it includes a detailed input manual and sample output, etc. (author). 55 refs.

  14. Preparation of data relevant to ''Equivalent Uniform Burnup'' and Equivalent Initial Enrichment'' for burnup credit evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Nomura, Yasushi; Okuno, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Murazaki, Minoru [Tokyo Nuclear Service Inc., Tokyo (Japan)

    2001-11-01

    Based on the PWR spent fuel composition data measured at JAERI, two kinds of simplified methods such as ''Equivalent Uniform Burnup'' and ''Equivalent Initial Enrichment'' have been introduced. And relevant evaluation curves have been prepared for criticality safety evaluation of spent fuel storage pool and transport casks, taking burnup of spent fuel into consideration. These simplified methods can be used to obtain an effective neutron multiplication factor for a spent fuel storage/transportation system by using the ORIGEN2.1 burnup code and the KENO-Va criticality code without considering axial burnup profile in spent fuel and other various factors introducing calculated errors. ''Equivalent Uniform Burnup'' is set up for its criticality analysis to be reactivity equivalent with the detailed analysis, in which the experimentally obtained isotopic composition together with a typical axial burnup profile and various factors such as irradiation history are considered on the conservative side. On the other hand, Equivalent Initial Enrichment'' is set up for its criticality analysis to be reactivity equivalent with the detailed analysis such as above when it is used in the so called fresh fuel assumption. (author)

  15. RAPID program to predict radial power and burnup distribution of UO{sub 2} fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Song, Jae Sung; Bang, Je Gun; Kim, Dae Ho [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-02-01

    Due to the radial variation of the neutron flux and its energy spectrum inside UO{sub 2} fuel, the fission density and fissile isotope production rates are varied radially in the pellet, and it becomes necessary to know the accurate radial power and burnup variation to predict the high burnup fuel behavior such as rim effects. Therefore, to predict the radial distribution of power, burnup and fissionable nuclide densities in the pellet with the burnup and U-235 enrichment, RAPID(RAdial power and burnup Prediction by following fissile Isotope Distribution in the pellet) program was developed. It considers the specific radial variation of the neutron reaction of the nuclides while the constant radial variation of neutron reaction except neutron absorption of U-238 regardless of the nuclides, the burnup and U-235 enrichment is assumed in TUBRNP model which is recognized as the one of the most reliable models. Therefore, it is expected that RAPID may be more accurate than TUBRNP, specially at high burnup region. RAPID is based upon and validated by the detailed reactor physics code, HELIOS which is one of few codes that can calculates the radial variations of the nuclides inside the pellet. Comparison of RAPID prediction with the measured data of the irradiated fuels showed very good agreement. RAPID can be used to calculate the local variations of the fissionable nuclide concentrations as well as the local power and burnup inside that pellet as a function of the burnup up to 10 w/o U-235 enrichment and 150 MWD/kgU burnup under the LWR environment. (author). 8 refs., 50 figs., 1 tab.

  16. Development of burnup dependent fuel rod model in COBRA-TF

    Science.gov (United States)

    Yilmaz, Mine Ozdemir

    The purpose of this research was to develop a burnup dependent fuel thermal conductivity model within Pennsylvania State University, Reactor Dynamics and Fuel Management Group (RDFMG) version of the subchannel thermal-hydraulics code COBRA-TF (CTF). The model takes into account first, the degradation of fuel thermal conductivity with high burnup; and second, the fuel thermal conductivity dependence on the Gadolinium content for both UO2 and MOX fuel rods. The modified Nuclear Fuel Industries (NFI) model for UO2 fuel rods and Duriez/Modified NFI Model for MOX fuel rods were incorporated into CTF and fuel centerline predictions were compared against Halden experimental test data and FRAPCON-3.4 predictions to validate the burnup dependent fuel thermal conductivity model in CTF. Experimental test cases from Halden reactor fuel rods for UO2 fuel rods at Beginning of Life (BOL), through lifetime without Gd2O3 and through lifetime with Gd 2O3 and a MOX fuel rod were simulated with CTF. Since test fuel rod and FRAPCON-3.4 results were based on single rod measurements, CTF was run for a single fuel rod surrounded with a single channel configuration. Input decks for CTF were developed for one fuel rod located at the center of a subchannel (rod-centered subchannel approach). Fuel centerline temperatures predicted by CTF were compared against the measurements from Halden experimental test data and the predictions from FRAPCON-3.4. After implementing the new fuel thermal conductivity model in CTF and validating the model with experimental data, CTF model was applied to steady state and transient calculations. 4x4 PWR fuel bundle configuration from Purdue MOX benchmark was used to apply the new model for steady state and transient calculations. First, one of each high burnup UO2 and MOX fuel rods from 4x4 matrix were selected to carry out single fuel rod calculations and fuel centerline temperatures predicted by CTF/TORT-TD were compared against CTF /TORT-TD /FRAPTRAN

  17. Boron coating on boron nitride coated nuclear fuels by chemical vapor deposition

    Science.gov (United States)

    Durmazuçar, Hasan H.; Gündüz, Güngör

    2000-12-01

    Uranium dioxide-only and uranium dioxide-gadolinium oxide (5% and 10%) ceramic nuclear fuel pellets which were already coated with boron nitride were coated with thin boron layer by chemical vapor deposition to increase the burn-up efficiency of the fuel during reactor operation. Coating was accomplished from the reaction of boron trichloride with hydrogen at 1250 K in a tube furnace, and then sintering at 1400 and 1525 K. The deposited boron was identified by infrared spectrum. The morphology of the coating was studied by using scanning electron microscope. The plate, grainy and string (fiber)-like boron structures were observed.

  18. The burnup dependence of light water reactor spent fuel oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, B.D.

    1998-07-01

    Over the temperature range of interest for dry storage or for placement of spent fuel in a permanent repository under the conditions now being considered, UO{sub 2} is thermodynamically unstable with respect to oxidation to higher oxides. The multiple valence states of uranium allow for the accommodation of interstitial oxygen atoms in the fuel matrix. A variety of stoichiometric and nonstoichiometric phases is therefore possible as the fuel oxidizers from UO{sub 2} to higher oxides. The oxidation of UO{sub 2} has been studied extensively for over 40 years. It has been shown that spent fuel and unirradiated UO{sub 2} oxidize via different mechanisms and at different rates. The oxidation of LWR spent fuel from UO{sub 2} to UO{sub 2.4} was studied previously and is reasonably well understood. The study presented here was initiated to determine the mechanism and rate of oxidation from UO{sub 2.4} to higher oxides. During the early stages of this work, a large variability in the oxidation behavior of samples oxidized under nearly identical conditions was found. Based on previous work on the effect of dopants on UO{sub 2} oxidation and this initial variability, it was hypothesized that the substitution of fission product and actinide impurities for uranium atoms in the spent fuel matrix was the cause of the variable oxidation behavior. Since the impurity concentration is roughly proportional to the burnup of a specimen, the oxidation behavior of spent fuel was expected to be a function of both temperature and burnup. This report (1) summarizes the previous oxidation work for both unirradiated UO{sub 2} and spent fuel (Section 2.2) and presents the theoretical basis for the burnup (i.e., impurity concentration) dependence of the rate of oxidation (Sections 2.3, 2.4, and 2.5), (2) describes the experimental approach (Section 3) and results (Section 4) for the current oxidation tests on spent fuel, and (3) establishes a simple model to determine the activation energies

  19. Mechanical Fatigue Testing of High Burnup Fuel for Transportation Applications

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-05-01

    This report describes testing designed to determine the ability of high burnup (HBU) (>45 GWd/MTU) spent fuel to maintain its integrity under normal conditions of transportation. An innovative system, Cyclic Integrated Reversible-bending Fatigue Tester (CIRFT), has been developed at Oak Ridge National Laboratory (ORNL) to test and evaluate the mechanical behavior of spent nuclear fuel (SNF) under conditions relevant to storage and transportation. The CIRFT system is composed of a U-frame equipped with load cells for imposing the pure bending loads on the SNF rod test specimen and measuring the in-situ curvature of the fuel rod during bending using a set up with three linear variable differential transformers (LVDTs).

  20. Manufacturing Data Uncertainties Propagation Method in Burn-Up Problems

    Directory of Open Access Journals (Sweden)

    Thomas Frosio

    2017-01-01

    Full Text Available A nuclear data-based uncertainty propagation methodology is extended to enable propagation of manufacturing/technological data (TD uncertainties in a burn-up calculation problem, taking into account correlation terms between Boltzmann and Bateman terms. The methodology is applied to reactivity and power distributions in a Material Testing Reactor benchmark. Due to the inherent statistical behavior of manufacturing tolerances, Monte Carlo sampling method is used for determining output perturbations on integral quantities. A global sensitivity analysis (GSA is performed for each manufacturing parameter and allows identifying and ranking the influential parameters whose tolerances need to be better controlled. We show that the overall impact of some TD uncertainties, such as uranium enrichment, or fuel plate thickness, on the reactivity is negligible because the different core areas induce compensating effects on the global quantity. However, local quantities, such as power distributions, are strongly impacted by TD uncertainty propagations. For isotopic concentrations, no clear trends appear on the results.

  1. Gallium nitride optoelectronic devices

    Science.gov (United States)

    Chu, T. L.; Chu, S. S.

    1972-01-01

    The growth of bulk gallium nitride crystals was achieved by the ammonolysis of gallium monochloride. Gallium nitride single crystals up to 2.5 x 0.5 cm in size were produced. The crystals are suitable as substrates for the epitaxial growth of gallium nitride. The epitaxial growth of gallium nitride on sapphire substrates with main faces of (0001) and (1T02) orientations was achieved by the ammonolysis of gallium monochloride in a gas flow system. The grown layers had electron concentrations in the range of 1 to 3 x 10 to the 19th power/cu cm and Hall mobilities in the range of 50 to 100 sq cm/v/sec at room temperature.

  2. Boron Nitride Nanotubes

    Science.gov (United States)

    Smith, Michael W. (Inventor); Jordan, Kevin (Inventor); Park, Cheol (Inventor)

    2012-01-01

    Boron nitride nanotubes are prepared by a process which includes: (a) creating a source of boron vapor; (b) mixing the boron vapor with nitrogen gas so that a mixture of boron vapor and nitrogen gas is present at a nucleation site, which is a surface, the nitrogen gas being provided at a pressure elevated above atmospheric, e.g., from greater than about 2 atmospheres up to about 250 atmospheres; and (c) harvesting boron nitride nanotubes, which are formed at the nucleation site.

  3. Boron nitride composites

    Science.gov (United States)

    Kuntz, Joshua D.; Ellsworth, German F.; Swenson, Fritz J.; Allen, Patrick G.

    2016-02-16

    According to one embodiment, a composite product includes hexagonal boron nitride (hBN), and a plurality of cubic boron nitride (cBN) particles, wherein the plurality of cBN particles are dispersed in a matrix of the hBN. According to another embodiment, a composite product includes a plurality of cBN particles, and one or more borate-containing binders.

  4. Boron nitride composites

    Energy Technology Data Exchange (ETDEWEB)

    Kuntz, Joshua D.; Ellsworth, German F.; Swenson, Fritz J.; Allen, Patrick G.

    2016-02-16

    According to one embodiment, a composite product includes hexagonal boron nitride (hBN), and a plurality of cubic boron nitride (cBN) particles, wherein the plurality of cBN particles are dispersed in a matrix of the hBN. According to another embodiment, a composite product includes a plurality of cBN particles, and one or more borate-containing binders.

  5. Burnup measurements on spent fuel elements of the RP-10 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vela Mora, Mariano; Gallardo Padilla, Alberto; Palomino, Jose Luis Castro, E-mail: mvela@ipen.gob.p [Instituto Peruano de Energia Nuclear (IPEN/Peru), Lima (Peru). Grupo de Calculo, Analisis y Seguridad de Reactores; Terremoto, Luis Antonio Albiac, E-mail: laaterre@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    This work describes the measurement, using nondestructive gamma-ray spectroscopy, of the average burnup attained by Material Testing Reactor (MTR) fuel elements irradiated in the RP-10 research reactor. Measurements were performed at the reactor storage pool area using {sup 137}Cs as the only burnup monitor, even for spent fuel elements with cooling times much shorter than two years. The experimental apparatus was previously calibrated in efficiency to obtain absolute average burnup values, which were compared against corresponding ones furnished by reactor physics calculations. The mean deviation between both values amounts to 6%. (author)

  6. Actinide-only and full burn-up credit in criticality assessment of RBMK-1500 spent nuclear fuel storage cask using axial burn-up profile

    Energy Technology Data Exchange (ETDEWEB)

    Barkauskas, V., E-mail: vytenis.barkauskas@ftmc.lt; Plukiene, R., E-mail: rita.plukiene@ftmc.lt; Plukis, A., E-mail: arturas.plukis@ftmc.lt

    2016-10-15

    Highlights: • RBMK-1500 fuel burn-up impact on k{sub eff} in the SNF cask was calculated using SCALE 6.1. • Positive end effect was noticed at certain burn-up for the RBMK-1500 spent nuclear fuel. • The non-uniform uranium depletion is responsible for the end effect in RBMK-1500 SNF. • k{sub eff} in the SNF cask does not exceed a value of 0.95 which is set in the safety requirements. - Abstract: Safe long-term storage of spent nuclear fuel (SNF) is one of the main issues in the field of nuclear safety. Burn-up credit application in criticality analysis of SNF reduces conservatism of usually used fresh fuel assumption and implies a positive economic impact for the SNF storage. Criticality calculations of spent nuclear fuel in the CONSTOR® RBMK-1500/M2 cask were performed using pre-generated ORIGEN-ARP spent nuclear fuel composition libraries, and the results of the RBMK-1500 burn-up credit impact on the effective neutron multiplication factor (k{sub eff}) have been obtained and are presented in the paper. SCALE 6.1 code package with the STARBUCKS burn-up credit evaluation tool was used for modeling. Pre-generated ARP (Automatic Rapid Processing) crosssection libraries based on ENDF/B-VII cross section library were used for fast burn-up inventory modeling. Different conditions in the SNF cask were modeled: 2.0% and 2.8% initial enrichment fuel of various burn-up and water density inside cavities of the SNF cask. The fuel composition for the criticality analysis was chosen taking into account main actinides and most important fission products used in burn-up calculations. A significant positive end effect is noticed from 15 GWd/tU burn-up for 2.8% enrichment fuel and from 9 GWd/tU for 2.0% enrichment fuel applying the actinide-only approach. The obtained results may be applied in further evaluations of the RBMK type reactor SNF storage as well as help to optimize the SNF storage volume inside the CONSTOR® RBMK-1500/M2 cask without compromising criticality

  7. Nitrogen Availability Of Nitriding Atmosphere In Controlled Gas Nitriding Processes

    Directory of Open Access Journals (Sweden)

    Michalski J.

    2015-06-01

    Full Text Available Parameters which characterize the nitriding atmosphere in the gas nitriding process of steel are: the nitriding potential KN, ammonia dissociation rate α and nitrogen availabilitymN2. The article discusses the possibilities of utilization of the nitriding atmosphere’s nitrogen availability in the design of gas nitriding processes of alloyed steels in atmospheres derived from raw ammonia, raw ammonia diluted with pre-dissociated ammonia, with nitrogen, as well as with both nitrogen and pre-dissociated ammonia. The nitriding processes were accomplished in four series. The parameters selected in the particular processes were: process temperature (T, time (t, value of nitriding potential (KN, corresponding to known dissociation rate of the ammonia which dissociates during the nitriding process (α. Variable parameters were: nitrogen availability (mN2, composition of the ingoing atmosphere and flow rate of the ingoing atmosphere (FIn.

  8. Technical Development on Burn-up Credit for Spent LWR Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I.C.

    2001-12-26

    Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled ''Technical Development on Criticality Safety Management for Spent LWR Fuels.'' Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burnup and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report.

  9. Technical development on burn-up credit for spent LWR fuels

    Energy Technology Data Exchange (ETDEWEB)

    Nakahara, Yoshinori; Suyama, Kenya; Suzaki, Takenori [eds.] [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-10-01

    Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled 'Technical Development on Criticality Safety Management for Spent LWR Fuels'. Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burn-up and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report. (author)

  10. ThO{sub 2}-UO{sub 2} annular pins for high burnup fuels

    Energy Technology Data Exchange (ETDEWEB)

    Caner, Marc; Dugan, Edward T

    2000-06-01

    The main purpose of this work is to investigate the use of annular fuel pins (particularly pins containing thorium dioxide) for high burnup fuel. The following parameters were evaluated and compared between postulated mixed thorium-uranium dioxide standard and annular (9% void fraction) type fuel assemblies, as a function of burnup: the infinite multiplication factor, the uranium and plutonium isotopic compositions, the fuel temperature coefficient of reactivity and the conversion ratio. We used the SCALE-4.3 code system. The calculation method consisted in obtaining actinide and fission product number densities as functions of assembly burnup, by means of a 1-D transport calculation combined with a 0-D burnup calculation. These number densities were then used in a 3-D Monte Carlo code for obtaining k{sub {infinity}} from two-dimensional-symmetry 'snapshots'.

  11. Decay heat power of spent nuclear fuel of power reactors with high burnup at long-term storage

    Science.gov (United States)

    Ternovykh, Mikhail; Tikhomirov, Georgy; Saldikov, Ivan; Gerasimov, Alexander

    2017-09-01

    Decay heat power of actinides and fission products from spent nuclear fuel of power VVER-1000 type reactors at long-term storage is calculated. Two modes of storage are considered: mode in which single portion of actinides or fission products is loaded in storage facility, and mode in which actinides or fission products from spent fuel of one VVER reactor are added every year in storage facility during 30 years and then accumulated nuclides are stored without addition new nuclides. Two values of fuel burnup 40 and 70 MW·d/kg are considered for the mode of storage of single fuel unloading. For the mode of accumulation of spent fuel with subsequent storage, one value of burnup of 70 MW·d/kg is considered. Very long time of storage 105 years accepted in calculations allows to simulate final geological disposal of radioactive wastes. Heat power of fission products decreases quickly after 50-100 years of storage. The power of actinides decreases very slow. In passing from 40 to 70 MW·d/kg, power of actinides increases due to accumulation of higher fraction of 244Cm. These data are important in the back end of fuel cycle when improved cooling system of the storage facility will be required along with stronger radiation protection during storage, transportation and processing.

  12. Metal Nitrides for Plasmonic Applications

    DEFF Research Database (Denmark)

    Naik, Gururaj V.; Schroeder, Jeremy; Guler, Urcan;

    2012-01-01

    Metal nitrides as alternatives to metals such as gold could offer many advantages when used as plasmonic material. We show that transition metal nitrides can replace metals providing equally good optical performance for many plasmonic applications.......Metal nitrides as alternatives to metals such as gold could offer many advantages when used as plasmonic material. We show that transition metal nitrides can replace metals providing equally good optical performance for many plasmonic applications....

  13. Topical Report on Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1998-09-01

    The objective of this topical report is to present to the NRC for review and acceptance a methodology for using burnup credit in the design of criticality control systems for PWR spent fuel transportation packages, while maintaining the criticality safety margins and related requirements of 10 CFR Part 71 and 72. The proposed methodology consists of five major steps as summarized below: (1) Validate a computer code system to calculate isotopic concentrations in SNF created during burnup in the reactor core and subsequent decay. (2) Validate a computer code system to predict the subcritical multiplication factor, keff, of a spent nuclear fuel package. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). and (5) Verify that SNF assemblies meet the package loading criteria and confirm proper fuel assembly selection prior to loading. (This step is required but the details are outside the scope of this topical report.) When reviewed and accepted by the NRC, this topical report will serve as a criterion document for criticality control analysts and will provide steps for the use of actinide-only burnup credit in the design of criticality control systems. The NRC-accepted burnup credit methodology will be used by commercial SNF storage and transportation package designers. Design-specific burnup credit criticality analyses will be defined, developed, and documented in the Safety Analysis Report (SAR) for each specific storage or transportation package that uses burnup credit. These SARs will then be submitted to the NRC for review and approval. This topical report is expected to be referenced in a number of storage and transportation cask applications to be submitted by commercial cask and canister designers to the NRC. Therefore, NRC acceptance of this topical report will result in increased efficiency of the

  14. Burnup analysis of the VVER-1000 reactor using thorium-based fuel

    Energy Technology Data Exchange (ETDEWEB)

    Korkmaz, Mehmet E.; Agar, Osman; Bueyueker, Eylem [Karamanoglu Mehmetbey Univ., Karaman (Turkey). Faculty of Kamil Ozdag Science

    2014-12-15

    This paper aims to investigate {sup 232}Th/{sup 233}U fuel cycles in a VVER-1000 reactor through calculation by computer. The 3D core geometry of VVER-1000 system was designed using the Serpent Monte Carlo 1.1.19 Code. The Serpent Code using parallel programming interface (Message Passing Interface-MPI), was run on a workstation with 12-core and 48 GB RAM. {sup 232}Th/{sup 235}U/{sup 238}U oxide mixture was considered as fuel in the core, when the mass fraction of {sup 232}Th was increased as 0.05-0.1-0.2-0.3-0.4 respectively, the mass fraction of {sup 238}U equally was decreased. In the system, the calculations were made for 3 000 MW thermal power. For the burnup analyses, the core is assumed to deplete from initial fresh core up to a burnup of 16 MWd/kgU without refuelling considerations. In the burnup calculations, a burnup interval of 360 effective full power days (EFPDs) was defined. According to burnup, the mass changes of the {sup 232}Th, {sup 233}U, {sup 238}U, {sup 237}Np, {sup 239}Pu, {sup 241}Am and {sup 244}Cm were evaluated, and also flux and criticality of the system were calculated in dependence of the burnup rate.

  15. Nitride quantum light sources

    Science.gov (United States)

    Zhu, T.; Oliver, R. A.

    2016-02-01

    Prototype nitride quantum light sources, particularly single-photon emitters, have been successfully demonstrated, despite the challenges inherent in this complex materials system. The large band offsets available between different nitride alloys have allowed device operation at easily accessible temperatures. A wide range of approaches has been explored: not only self-assembled quantum dot growth but also lithographic methods for site-controlled nanostructure formation. All these approaches face common challenges, particularly strong background signals which contaminate the single-photon stream and excessive spectral diffusion of the quantum dot emission wavelength. If these challenges can be successfully overcome, then ongoing rapid progress in the conventional III-V semiconductors provides a roadmap for future progress in the nitrides.

  16. Technical Data to Justify Full Burnup Credit in Criticality Safety Licensing Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Enercon Services, Inc.

    2011-03-14

    Enercon Services, Inc. (ENERCON) was requested under Task Order No.2 to identify scientific and technical data needed to benchmark and justify Full Burnup Credit, which adds 16 fission products and 4 minor actinides1 to Actinide-Only burnup credit. The historical perspective for Full Burnup Credit is discussed, and interviews of organizations participating in burnup credit activities are summarized as a basis for identifying additional data needs and making recommendation. Input from burnup credit participants representing two segments of the commercial nuclear industry is provided. First, the Electric Power Research Institute (EPRI) has been very active in the development of Full Burnup Credit, representing the interests of nuclear utilities in achieving capacity gains for storage and transport casks. EPRI and its utility customers are interested in a swift resolution of the validation issues that are delaying the implementation of Full Burnup Credit [EPRI 2010b]. Second, used nuclear fuel storage and transportation Cask Vendors favor improving burnup credit beyond Actinide-Only burnup credit, although their discussion of specific burnup credit achievements and data needs was limited citing business sensitive and technical proprietary concerns. While Cask Vendor proprietary items are not specifically identified in this report, the needs of all nuclear industry participants are reflected in the conclusions and recommendations of this report. In addition, Oak Ridge National Laboratory (ORNL) and Sandia National Laboratory (SNL) were interviewed for their input into additional data needs to achieve Full Burnup Credit. ORNL was very open to discussions of Full Burnup Credit, with several telecoms and a visit by ENERCON to ORNL. For many years, ORNL has provided extensive support to the NRC regarding burnup credit in all of its forms. Discussions with ORNL focused on potential resolutions to the validation issues for the use of fission products. SNL was helpful in

  17. Coolant Density and Control Blade History Effects in Extended BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian J [ORNL; Marshall, William BJ J [ORNL; Bowman, Stephen M [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Martinez-Gonzalez, Jesus S [ORNL

    2015-01-01

    Oak Ridge National Laboratory and the US Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (keff) calculations and predicted spent fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date investigating some aspects of extended BUC. (The technical basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper.) Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC: (1) the effect of axial void profile and (2) the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of a modern operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. Although a single cycle does not provide complete data, the data obtained are sufficient to determine the primary effects and to identify conservative modeling approaches. These data were used in a study of the effect of axial void profile. The first stage of the study was determination of the necessary moderator density temporal fidelity in depletion modeling. After the required temporal fidelity was established, multiple void profiles were used to examine the effect on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied was control blade history. Control blades are inserted in

  18. High Burn-Up Spent Nuclear Fuel Vibration Integrity Study

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jiang, Hao [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Rob L [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John M [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

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

  19. Topical report on actinide-only burnup credit for PWR spent nuclear fuel packages. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1997-04-01

    A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, k{sub eff}, of a spent nuclear fuel package. Fifty-seven UO{sub 2}, UO{sub 2}/Gd{sub 2}O{sub 3}, and UO{sub 2}/PuO{sub 2} critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k{sub eff} (which can be a function of the trending parameters) such that the biased k{sub eff}, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package

  20. Tritium release from EXOTIC-7 orthosilicate pebbles. Effect of burnup and contact with beryllium during irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Scaffidi-Argentina, F.; Werle, H. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik

    1998-03-01

    EXOTIC-7 was the first in-pile test with {sup 6}Li-enriched (50%) lithium orthosilicate (Li{sub 4}SiO{sub 4}) pebbles and with DEMO representative Li-burnup. Post irradiation examinations of the Li{sub 4}SiO{sub 4} have been performed at the Forschungszentrum Karlsruhe (FZK), mainly to investigate the tritium release kinetics as well as the effect of Li-burnup and/or contact with beryllium during irradiation. The release rate of Li{sub 4}SiO{sub 4} from pure Li{sub 4}SiO{sub 4} bed of capsule 28.1-1 is characterized by a broad main peak at about 400degC and by a smaller peak at about 800degC, and that from the mixed beds of capsule 28.2 and 26.2-1 shows again these two peaks, but most of the tritium is now released from the 800degC peak. This shift of release from low to high temperature may be due to the higher Li-burnup and/or due to contact with Be during irradiation. Due to the very difficult interpretation of the in-situ tritium release data, residence times have been estimated on the basis of the out-of-pile tests. The residence time for Li{sub 4}SiO{sub 4} from caps. 28.1-1 irradiated at 10% Li-burnup agrees quite well with that of the same material irradiated at Li-burnup lower than 3% in the EXOTIC-6 experiment. In spite of the observed shift in the release peaks from low to high temperature, also the residence time for Li{sub 4}SiO{sub 4} from caps. 26.2-1 irradiated at 13% Li-burnup agrees quite well with the data from EXOTIC-6 experiment. On the other hand, the residence time for Li{sub 4}SiO{sub 4} from caps. 28.2 (Li-burnup 18%) is about a factor 1.7-3.8 higher than that for caps. 26.2-1. Based on these data on can conclude that up to 13% Li-burnup neither the contact with beryllium nor the Li-burnup have a detrimental effect on the tritium release of Li{sub 4}SiO{sub 4} pebbles, but at 18% Li-burnup the residence time is increased by about a factor three. (J.P.N.)

  1. Assessment of the uncertainties of MULTICELL calculations by the OECD NEA UAM PWR pin cell burnup benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Kereszturi, Andras [Hungarian Academy of Sciences, Budapest (Hungary). Centre for Energy Research; Panka, Istvan

    2015-09-15

    Defining precisely the burnup of the nuclear fuel is important from the point of view of core design calculations, safety analyses, criticality calculations (e.g. burnup credit calculations), etc. This paper deals with the uncertainties of MULTICELL calculations obtained by the solution of the OECD NEA UAM PWR pin cell burnup benchmark. In this assessment Monte-Carlo type statistical analyses are applied and the energy dependent covariance matrices of the cross-sections are taken into account. Additionally, the impact of the uncertainties of the fission yields is also considered. The target quantities are the burnup dependent uncertainties of the infinite multiplication factor, the two-group cross-sections, the reaction rates and the number densities of some isotopes up to the burnup of 60 MWd/kgU. In the paper the burnup dependent tendencies of the corresponding uncertainties and their sources are analyzed.

  2. Pair distribution functions of silicon/silicon nitride interfaces

    Science.gov (United States)

    Cao, Deng; Bachlechner, Martina E.

    2006-03-01

    Using molecular dynamics simulations, we investigate different mechanical and structural properties of the silicon/silicon nitride interface. One way to characterize the structure as tensile strain is applied parallel to the interface is to calculate pair distribution functions for specific atom types. The pair distribution function gives the probability of finding a pair of atoms a distance r apart, relative to the probability expected for a completely random distribution at the same density. The pair distribution functions for bulk silicon nitride reflect the fracture of the silicon nitride film at about 8 % and the fact that the centerpiece of the silicon nitride film returns to its original structure after fracture. The pair distribution functions for interface silicon atoms reveal the formation of bonds for originally unbound atom pairs, which is indicative of the interstitial-vacancy defect that causes failure in silicon.

  3. Bond Angles in the Crystalline Silicon/Silicon Nitride Interface

    Science.gov (United States)

    Leonard, Robert H.; Bachlechner, Martina E.

    2006-03-01

    Silicon nitride deposited on a silicon substrate has major applications in both dielectric layers in microelectronics and as antireflection and passivation coatings in photovoltaic applications. Molecular dynamic simulations are performed to investigate the influence of temperature and rate of externally applied strain on the structural and mechanical properties of the silicon/silicon nitride interface. Bond-angles between various atom types in the system are used to find and understand more about the mechanisms leading to the failure of the crystal. Ideally in crystalline silicon nitride, bond angles of 109.5 occur when a silicon atom is at the vertex and 120 angles occur when a nitrogen atom is at the vertex. The comparison of the calculated angles to the ideal values give information on the mechanisms of failure in silicon/silicon nitride system.

  4. New high burnup fuel models for NRC`s licensing audit code, FRAPCON

    Energy Technology Data Exchange (ETDEWEB)

    Lanning, D.D.; Beyer, C.E.; Painter, C.L. [Pacific Northwest Laboratory, Richland, WA (United States)

    1996-03-01

    Fuel behavior models have recently been updated within the U.S. Nuclear Regulatory Commission steady-state FRAPCON code used for auditing of fuel vendor/utility-codes and analyses. These modeling updates have concentrated on providing a best estimate prediction of steady-state fuel behavior up to the maximum burnup level s of current data (60 to 65 GWd/MTU rod-average). A decade has passed since these models were last updated. Currently, some U.S. utilities and fuel vendors are requesting approval for rod-average burnups greater than 60 GWd/MTU; however, until these recent updates the NRC did not have valid fuel performance models at these higher burnup levels. Pacific Northwest Laboratory (PNL) has reviewed 15 separate effects models within the FRAPCON fuel performance code (References 1 and 2) and identified nine models that needed updating for improved prediction of fuel behavior at high burnup levels. The six separate effects models not updated were the cladding thermal properties, cladding thermal expansion, cladding creepdown, fuel specific heat, fuel thermal expansion and open gap conductance. Comparison of these models to the currently available data indicates that these models still adequately predict the data within data uncertainties. The nine models identified as needing improvement for predicting high-burnup behavior are fission gas release (FGR), fuel thermal conductivity (accounting for both high burnup effects and burnable poison additions), fuel swelling, fuel relocation, radial power distribution, fuel-cladding contact gap conductance, cladding corrosion, cladding mechanical properties and cladding axial growth. Each of the updated models will be described in the following sections and the model predictions will be compared to currently available high burnup data.

  5. Summary of high burnup fuel issues and NRC`s plan of action

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, R.O.

    1997-01-01

    For the past two years the Office of Nuclear Regulatory Research has concentrated mostly on the so-called reactivity-initiated accidents -- the RIAs -- in this session of the Water Reactor Safety Information Meeting, but this year there is a more varied agenda. RIAs are, of course, not the only events of interest for reactor safety that are affected by extended burnup operation. Their has now been enough time to consider a range of technical issues that arise at high burnup, and a list of such issues being addressed in their research program is given here. (1) High burnup capability of the steady-state code (FRAPCON) used for licensing audit calculations. (2) General capability (including high burnup) of the transient code (FRAPTRAN) used for special studies. (3) Adequacy at high burnup of fuel damage criteria used in regulation for reactivity accidents. (4) Adequacy at high burnup of models and fuel related criteria used in regulation for loss-of-coolant accidents (LOCAs). (5) Effect of high burnup on fuel system damage during normal operation, including control rod insertion problems. A distinction is made between technical issues, which may or may not have direct licensing impacts, and licensing issues. The RIAs became a licensing issue when the French test in CABRI showed that cladding failures could occur at fuel enthalpies much lower than a value currently used in licensing. Fuel assembly distortion became a licensing issue when control rod insertion was affected in some operating plants. In this presentation, these technical issues will be described and the NRC`s plan of action to address them will be discussed.

  6. Metal surface nitriding by laser induced plasma

    Science.gov (United States)

    Thomann, A. L.; Boulmer-Leborgne, C.; Andreazza-Vignolle, C.; Andreazza, P.; Hermann, J.; Blondiaux, G.

    1996-10-01

    We study a nitriding technique of metals by means of laser induced plasma. The synthesized layers are composed of a nitrogen concentration gradient over several μm depth, and are expected to be useful for tribological applications with no adhesion problem. The nitriding method is tested on the synthesis of titanium nitride which is a well-known compound, obtained at present by many deposition and diffusion techniques. In the method of interest, a laser beam is focused on a titanium target in a nitrogen atmosphere, leading to the creation of a plasma over the metal surface. In order to understand the layer formation, it is necessary to characterize the plasma as well as the surface that it has been in contact with. Progressive nitrogen incorporation in the titanium lattice and TiN synthesis are studied by characterizing samples prepared with increasing laser shot number (100-4000). The role of the laser wavelength is also inspected by comparing layers obtained with two kinds of pulsed lasers: a transversal-excited-atmospheric-pressure-CO2 laser (λ=10.6 μm) and a XeCl excimer laser (λ=308 nm). Simulations of the target temperature rise under laser irradiation are performed, which evidence differences in the initial laser/material interaction (material heated thickness, heating time duration, etc.) depending on the laser features (wavelength and pulse time duration). Results from plasma characterization also point out that the plasma composition and propagation mode depend on the laser wavelength. Correlation of these results with those obtained from layer analyses shows at first the important role played by the plasma in the nitrogen incorporation. Its presence is necessary and allows N2 dissociation and a better energy coupling with the target. Second, it appears that the nitrogen diffusion governs the nitriding process. The study of the metal nitriding efficiency, depending on the laser used, allows us to explain the differences observed in the layer features

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

    Science.gov (United States)

    Su'ud, Zaki; Sekimoto, H.

    2014-09-01

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

  8. Plutonium and Minor Actinides Recycling in Standard BWR using Equilibrium Burnup Model

    Directory of Open Access Journals (Sweden)

    Abdul Waris

    2008-03-01

    Full Text Available Plutonium (Pu and minor actinides (MA recycling in standard BWR with equilibrium burnup model has been studied. We considered the equilibrium burnup model as a simple time independent burnup method, which can manage all possible produced nuclides in any nuclear system. The equilibrium burnup code was bundled with a SRAC cell-calculation code to become a coupled cell-burnup calculation code system. The results show that the uranium enrichment for the criticality of the reactor, the amount of loaded fuel and the required natural uranium supply per year decrease for the Pu recycling and even much lower for the Pu & MA recycling case compared to those of the standard once-through BWR case. The neutron spectra become harder with the increasing number of recycled heavy nuclides in the reactor core. The total fissile rises from 4.77% of the total nuclides number density in the reactor core for the standard once-through BWR case to 6.64% and 6.72% for the Plutonium recycling case and the Pu & MA recycling case, respectively. The two later data may become the main basis why the required uranium enrichment declines and consequently diminishes the annual loaded fuel and the required natural uranium supply. All these facts demonstrate the advantage of plutonium and minor actinides recycling in BWR.

  9. Model for evolution of grain size in the rim region of high burnup UO2 fuel

    Science.gov (United States)

    Xiao, Hongxing; Long, Chongsheng; Chen, Hongsheng

    2016-04-01

    The restructuring process of the high burnup structure (HBS) formation in UO2 fuel results in sub-micron size grains that accelerate the fission gas swelling, which will raise some concern over the safety of extended the nuclear fuel operation life in the reactor. A mechanistic and engineering model for evolution of grain size in the rim region of high burnup UO2 fuel based on the experimental observations of the HBS in the literature is presented. The model takes into account dislocations evolution under irradiation and the grain subdivision occur successively at increasing local burnup. It is assumed that the original driving force for subdivision of grain in the HBS of UO2 fuel is the production and accumulation of dislocation loops during irradiation. The dislocation loops can also be annealed through thermal diffusion when the temperature is high enough. The capability of this model is validated by the comparison with the experimental data of temperature threshold of subdivision, dislocation density and sub-grain size as a function of local burnup. It is shown that the calculated results of the dislocation density and subdivided grain size as a function of local burnup are in good agreement with the experimental results.

  10. Analysis of plasma nitrided steels

    Science.gov (United States)

    Salik, J.; Ferrante, J.; Honecy, F.; Hoffman, R., Jr.

    1987-01-01

    The analysis of plasma nitrided steels can be divided to two main categories - structural and chemical. Structural analysis can provide information not only on the hardening mechanisms but also on the fundamental processes involved. Chemical analysis can be used to study the kinetics for the nitriding process and its mechanisms. In this paper preliminary results obtained by several techniques of both categories are presented and the applicability of those techniques to the analysis of plasma-nitrided steels is discussed.

  11. Boron nitride encapsulated graphene infrared emitters

    Energy Technology Data Exchange (ETDEWEB)

    Barnard, H. R.; Zossimova, E.; Mahlmeister, N. H.; Lawton, L. M.; Luxmoore, I. J.; Nash, G. R., E-mail: g.r.nash@exeter.ac.uk [College of Engineering, Mathematics and Physical Sciences, University of Exeter, Exeter EX4 4QF (United Kingdom)

    2016-03-28

    The spatial and spectral characteristics of mid-infrared thermal emission from devices containing a large area multilayer graphene layer, encapsulated using hexagonal boron nitride, have been investigated. The devices were run continuously in air for over 1000 h, with the emission spectrum covering the absorption bands of many important gases. An approximate solution to the heat equation was used to simulate the measured emission profile across the devices yielding an estimated value of the characteristic length, which defines the exponential rise/fall of the temperature profile across the device, of 40 μm. This is much larger than values obtained in smaller exfoliated graphene devices and reflects the device geometry, and the increase in lateral heat conduction within the devices due to the multilayer graphene and boron nitride layers.

  12. Review of Halden Reactor Project high burnup fuel data that can be used in safety analyses

    Energy Technology Data Exchange (ETDEWEB)

    Wiesenack, W. [OECD Halden Reactor Project (Norway)

    1996-03-01

    The fuels and materials testing programmes carried out at the OECD Halden Reactor Project are aimed at providing data in support of a mechanistic understanding of phenomena, especially as related to high burnup fuel. The investigations are focused on identifying long term property changes, and irradiation techniques and instrumentation have been developed over the years which enable to assess fuel behaviour and properties in-pile. The fuel-cladding gap has an influence on both thermal and mechanical behaviour. Improved gap conductance due to gap closure at high exposure is observed even in the case of a strong contamination with released fission gas. On the other hand, pellet-cladding mechanical interaction, which is measured with cladding elongation detectors and diameter gauges, is re-established after a phase with less interaction and is increasing. These developments are exemplified with data showing changes of fuel temperature, hydraulic diameter and cladding elongation with burnup. Fuel swelling and cladding primary and secondary creep have been successfully measured in-pile. They provide data for, e.g., the possible cladding lift-off to be accounted for at high burnup. Fuel conductivity degradation is observed as a gradual temperature increase with burnup. This affects stored heat, fission gas release and temperature dependent fuel behaviour in general. The Halden Project`s data base on fission gas release shows that the phenomenon is associated with an accumulation of gas atoms at the grain boundaries to a critical concentration before appreciable release occurs. This is accompanied by an increase of the surface-to-volume ratio measured in-pile in gas flow experiments. A typical observation at high burnup is also that a burst release of fission gas may occur during a power decrease. Gas flow and pressure equilibration experiments have shown that axial communication is severely restricted at high burnup.

  13. Impact investigation of reactor fuel operating parameters on reactivity for use in burnup credit applications

    Science.gov (United States)

    Sloma, Tanya Noel

    When representing the behavior of commercial spent nuclear fuel (SNF), credit is sought for the reduced reactivity associated with the net depletion of fissile isotopes and the creation of neutron-absorbing isotopes, a process that begins when a commercial nuclear reactor is first operated at power. Burnup credit accounts for the reduced reactivity potential of a fuel assembly and varies with the fuel burnup, cooling time, and the initial enrichment of fissile material in the fuel. With regard to long-term SNF disposal and transportation, tremendous benefits, such as increased capacity, flexibility of design and system operations, and reduced overall costs, provide an incentive to seek burnup credit for criticality safety evaluations. The Nuclear Regulatory Commission issued Interim Staff Guidance 8, Revision 2 in 2002, endorsing burnup credit of actinide composition changes only; credit due to actinides encompasses approximately 30% of exiting pressurized water reactor SNF inventory and could potentially be increased to 90% if fission product credit were accepted. However, one significant issue for utilizing full burnup credit, compensating for actinide and fission product composition changes, is establishing a set of depletion parameters that produce an adequately conservative representation of the fuel's isotopic inventory. Depletion parameters can have a significant effect on the isotopic inventory of the fuel, and thus the residual reactivity. This research seeks to quantify the reactivity impact on a system from dominant depletion parameters (i.e., fuel temperature, moderator density, burnable poison rod, burnable poison rod history, and soluble boron concentration). Bounding depletion parameters were developed by statistical evaluation of a database containing reactor operating histories. The database was generated from summary reports of commercial reactor criticality data. Through depletion calculations, utilizing the SCALE 6 code package, several light

  14. Preliminary Neutronic Design of High Burnup OTTO Cycle Pebble Bed Reactor

    Directory of Open Access Journals (Sweden)

    T. Setiadipura

    2015-04-01

    Full Text Available The pebble bed type High Temperature Gas-cooled Reactor (HTGR is among the interesting nuclear reactor designs in terms of safety and flexibility for co-generation applications. In addition, the strong inherent safety characteristics of the pebble bed reactor (PBR which is based on natural mechanisms improve the simplicity of the PBR design, in particular for the Once-Through-Then-Out (OTTO cycle PBR design. One of the important challenges of the OTTO cycle PBR design, and nuclear reactor design in general, is improving the nuclear fuel utilization which is shown by attaining a higher burnup value. This study performed a preliminary neutronic design study of a 200 MWt OTTO cycle PBR with high burnup while fulfilling the safety criteria of the PBR design.The safety criteria of the design was represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. The maximum burnup value was also limited by the tested maximum burnup value which maintained the integrity of the pebble fuel. Parametric surveys were performed to obtain the optimized parameters used in this study, which are the fuel enrichment, per-pebble heavy metal (HM loading, and the average axial speed of the fuel. An optimum design with burnup value of 131.1 MWd/Kg-HM was achieved in this study which is much higher compare to the burnup of the reference design HTR-MODUL and a previously proposed OTTO-cycle PBR design. This optimum design uses 17% U-235 enrichment with 4 g HM-loading per fuel pebble

  15. Sensitivity and parametric evaluations of significant aspects of burnup credit for PWR spent fuel packages

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    1996-05-01

    Spent fuel transportation and storage cask designs based on a burnup credit approach must consider issues that are not relevant in casks designed under a fresh-fuel loading assumption. For example, the spent fuel composition must be adequately characterized and the criticality analysis model can be complicated by the need to consider axial burnup variations. Parametric analyses are needed to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models must be evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. The purpose of this report is to describe analyses and evaluations performed in order to demonstrate the effect physical parameters and modeling assumptions have on the criticality analysis of spent fuel. The analyses in this report include determination and ranking of the most important actinides and fission products; study of the effect of various depletion scenarios on subsequent criticality calculations; establishment of trends in neutron multiplication as a function of fuel enrichment, burnup, cooling time- and a parametric and modeling evaluation of three-dimensional effects (e.g., axially varying burnup and temperature/density effects) in a conceptual cask design. The sensitivity and parametric evaluations were performed with the consideration of two different burnup credit approaches: (1) only actinides in the fuel are considered in the criticality analysis, and (2) both actinides and fission products are considered. Calculations described in this report were performed using the criticality and depletion sequences available in the SCALE code system and the SCALE 27-group burnup library. Although the results described herein do not constitute a validation of SCALE for use in spent fuel analysis, independent validation efforts have been completed and are described in other reports.

  16. Titanium Nitride Cermets

    Science.gov (United States)

    1952-07-01

    C ermets 7 Effect of Amount of Metal on Strength of TiN-Ni-Cr....26 Cerme ts S Effect of Amount of Metal on Strength of TiN-Co-Cr....27 Cermets 9...Figures 7 and 8. Titanium Nitride-Nickel-Chromium Cerme ts From Figure 7, it can be seen that 2900OF was the better firing temperature. The 20% metal

  17. Post Irradiation Examination Plan for High-Burnup Demonstration Project Sister Rods

    Energy Technology Data Exchange (ETDEWEB)

    Scaglione, John M [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Montgomery, Rose [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-04-01

    This test plan describes the experimental work to be implemented by the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) to characterize high burnup (HBU) spent nuclear fuel (SNF) in conjunction with the High Burnup Dry Storage Cask Research and Development Project and serves to coordinate and integrate the multi-year experimental program to collect and develop data regarding the continued storage and eventual transport of HBU (i.e., >45 GWd/MTU) SNF. The work scope involves the development, performance, technical integration, and oversight of measurements and collection of relevant data, guided by analyses and demonstration of need.

  18. Fuel burnup analysis for Thai research reactor by using MCNPX computer code

    Science.gov (United States)

    Sangkaew, S.; Angwongtrakool, T.; Srimok, B.

    2017-06-01

    This paper presents the fuel burnup analysis of the Thai research reactor (TRR-1/M1), TRIGA Mark-III, operated by Thailand Institute of Nuclear Technology (TINT) in Bangkok, Thailand. The modelling software used in this analysis is MCNPX (MCNP eXtended) version 2.6.0, a Fortran90 Monte Carlo radiation transport computer code. The analysis results will cover the core excess reactivity, neutron fluxes at the irradiation positions and neutron detector tubes, power distribution, fuel burnup, and fission products based on fuel cycle of first reactor core arrangement.

  19. Fuel burnup calculation of Ghana MNSR using ORIGEN2 and REBUS3 codes.

    Science.gov (United States)

    Abrefah, R G; Nyarko, B J B; Fletcher, J J; Akaho, E H K

    2013-10-01

    Ghana Research Reactor-1 core is to be converted from HEU fuel to LEU fuel in the near future and managing the spent nuclear fuel is very important. A fuel depletion analysis of the GHARR-1 core was performed using ORIGEN2 and REBUS3 codes to estimate the isotopic inventory at end-of-cycle in order to help in the design of an appropriate spent fuel cask. The results obtained for both codes were consistent for U-235 burnup weight percent and Pu-239 build up as a result of burnup.

  20. Progress of the RIA experiments with high burnup fuels and their evaluation in JAERI

    Energy Technology Data Exchange (ETDEWEB)

    Ishijima, Kiyomi; Fuketa, Toyoshi [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)

    1997-01-01

    Recent results obtained in the NSRR power burst experiments with high burnup PWR fuel rods are described and discussed in this paper. Data concerning test condition, transient records during pulse irradiation and post irradiation examination are described. Another high burnup PWR fuel rod failed in the test HBO-5 at the slightly higher energy deposition than that in the test HBO-1. The failure mechanism of the test HBO-5 is the same as that of the test HBO-1, that is, hydride-assisted PCMI. Some influence of the thermocouples welding on the failure behavior of the HBO-5 rod was observed.

  1. Plasmonic Titanium Nitride Nanostructures via Nitridation of Nanopatterned Titanium Dioxide

    DEFF Research Database (Denmark)

    Guler, Urcan; Zemlyanov, Dmitry; Kim, Jongbum

    2017-01-01

    Plasmonic titanium nitride nanostructures are obtained via nitridation of titanium dioxide. Nanoparticles acquired a cubic shape with sharper edges following the rock-salt crystalline structure of TiN. Lattice constant of the resulting TiN nanoparticles matched well with the tabulated data. Energ...

  2. Gallium nitride electronics

    Science.gov (United States)

    Rajan, Siddharth; Jena, Debdeep

    2013-07-01

    In the past two decades, there has been increasing research and industrial activity in the area of gallium nitride (GaN) electronics, stimulated first by the successful demonstration of GaN LEDs. While the promise of wide band gap semiconductors for power electronics was recognized many years before this by one of the contributors to this issue (J Baliga), the success in the area of LEDs acted as a catalyst. It set the field of GaN electronics in motion, and today the technology is improving the performance of several applications including RF cell phone base stations and military radar. GaN could also play a very important role in reducing worldwide energy consumption by enabling high efficiency compact power converters operating at high voltages and lower frequencies. While GaN electronics is a rapidly evolving area with active research worldwide, this special issue provides an opportunity to capture some of the great advances that have been made in the last 15 years. The issue begins with a section on epitaxy and processing, followed by an overview of high-frequency HEMTs, which have been the most commercially successful application of III-nitride electronics to date. This is followed by review and research articles on power-switching transistors, which are currently of great interest to the III-nitride community. A section of this issue is devoted to the reliability of III-nitride devices, an area that is of increasing significance as the research focus has moved from not just high performance but also production-worthiness and long-term usage of these devices. Finally, a group of papers on new and relatively less studied ideas for III-nitride electronics, such as interband tunneling, heterojunction bipolar transistors, and high-temperature electronics is included. These areas point to new areas of research and technological innovation going beyond the state of the art into the future. We hope that the breadth and quality of articles in this issue will make it

  3. FY14 Status Report: CIRFT Testing Results on High Burnup UNF

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-01

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

  4. Platinum nitride with fluorite structure

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Rong; Zhang, Xiao-Feng

    2005-01-31

    The mechanical stability of platinum nitride has been studied using first-principles calculations. By calculating the single-crystal elastic constants, we show that platinum nitride can be stabilized in the fluorite structure, in which the nitrogen atoms occupy all the tetrahedral interstitial sites of the metal lattice. The stability is attributed to the pseudogap effect from analysis of the electronic structure.

  5. Optical characterization of gallium nitride

    NARCIS (Netherlands)

    Kirilyuk, Victoria

    2002-01-01

    Group III-nitrides have been considered a promising system for semiconductor devices since a few decades, first for blue- and UV-light emitting diodes, later also for high-frequency/high-power applications. Due to the lack of native substrates, heteroepitaxially grown III-nitride layers are usually

  6. Electrochemical nitridation of metal surfaces

    Science.gov (United States)

    Wang, Heli; Turner, John A.

    2015-06-30

    Electrochemical nitridation of metals and the produced metals are disclosed. An exemplary method of electrochemical nitridation of metals comprises providing an electrochemical solution at low temperature. The method also comprises providing a three-electrode potentiostat system. The method also comprises stabilizing the three-electrode potentiostat system at open circuit potential. The method also comprises applying a cathodic potential to a metal.

  7. Dependence of heavy metal burnup on nuclear data libraries for fast reactors

    CERN Document Server

    Ohki, S

    2003-01-01

    Japan Nuclear Cycle Development Institute (JNC) is considering the highly burnt fuel as well as the recycling of minor actinide (MA) in the development of commercialized fast reactor cycle systems. Higher accuracy in burnup calculation is going to be required for higher mass plutonium isotopes ( sup 2 sup 4 sup 0 Pu, etc.) and MA nuclides. In the framework of research and development aiming at the validation and necessary improvements of fast reactor burnup calculation, we investigated the differences among the burnup calculation results with the major nuclear data libraries: JEF-2.2, ENDF/B-VI Release 5, JENDL-3.2, and JENDL-3.3. We focused on the heavy metal nuclides such as plutonium and MA in the central core region of a conventional sodium-cooled fast reactor. For main heavy metal nuclides ( sup 2 sup 3 sup 5 U, sup 2 sup 3 sup 8 U, sup 2 sup 3 sup 9 Pu, sup 2 sup 4 sup 0 Pu, and sup 2 sup 4 sup 1 Pu), number densities after 1-cycle burnup did not change over one or two percent. Library dependence was re...

  8. Development of an MCNP-tally based burnup code and validation through PWR benchmark exercises

    Energy Technology Data Exchange (ETDEWEB)

    El Bakkari, B. [ERSN-LMR, Department of physics, Faculty of Sciences P.O.Box 2121, Tetuan (Morocco)], E-mail: bakkari@gmail.com; El Bardouni, T.; Merroun, O.; El Younoussi, Ch.; Boulaich, Y. [ERSN-LMR, Department of physics, Faculty of Sciences P.O.Box 2121, Tetuan (Morocco); Chakir, E. [EPTN-LPMR, Faculty of Sciences Kenitra (Morocco)

    2009-05-15

    The aim of this study is to evaluate the capabilities of a newly developed burnup code called BUCAL1. The code provides the full capabilities of the Monte Carlo code MCNP5, through the use of the MCNP tally information. BUCAL1 uses the fourth order Runge Kutta method with the predictor-corrector approach as the integration method to determine the fuel composition at a desired burnup step. Validation of BUCAL1 was done by code vs. code comparison. Results of two different kinds of codes are employed. The first one is CASMO-4, a deterministic multi-group two-dimensional transport code. The second kind is MCODE and MOCUP, a link MCNP-ORIGEN codes. These codes use different burnup algorithms to solve the depletion equations system. Eigenvalue and isotope concentrations were compared for two PWR uranium and thorium benchmark exercises at cold (300 K) and hot (900 K) conditions, respectively. The eigenvalue comparison between BUCAL1 and the aforementioned two kinds of codes shows a good prediction of the systems'k-inf values during the entire burnup history, and the maximum difference is within 2%. The differences between the BUCAL1 isotope concentrations and the predictions of CASMO-4, MCODE and MOCUP are generally better, and only for a few sets of isotopes these differences exceed 10%.

  9. Radiochemical Assays of Irradiated VVER-440 Fuel for Use in Spent Fuel Burnup Credit Activities

    Energy Technology Data Exchange (ETDEWEB)

    Jardine, L J

    2005-04-25

    The objective of this spent fuel burnup credit work was to study and describe a VVER-440 reactor spent fuel assembly (FA) initial state before irradiation, its operational irradiation history and the resulting radionuclide distribution in the fuel assembly after irradiation. This work includes the following stages: (1) to pick out and select a specific spent (irradiated) FA for examination; (2) to describe the FA initial state before irradiation; (3) to describe the irradiation history, including thermal calculations; (4) to examine the burnup distribution of select radionuclides along the FA height and cross-section; (5) to examine the radionuclide distributions; (6) to determine the Kr-85 release into the plenum; (7) to select and prepare FA rod specimens for destructive examinations; (8) to determine the radionuclide compositions, isotope masses and burnup in the rod specimens; and (9) to analyze, document and process the results. The specific workscope included the destructive assay (DA) of spent fuel assembly rod segments with an {approx}38.5 MWd/KgU burnup from a single VVER-440 fuel assembly from the Novovorenezh reactor in Russia. Based on irradiation history criteria, four rods from the fuel assembly were selected and removed from the assembly for examination. Next, 8 sections were cut from the four rods and sent for destructive analysis of radionuclides by radiochemical analyses. The results were documented in a series of seven reports over a period of {approx}1 1/2 years.

  10. Reactivity effect of spent fuel depending on burn-up history

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, Takafumi [Nagoya Univ., Nagoya, Aichi (Japan); Suyama, Kenya; Nomura, Yasushi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Mochizuki, Hiroki [The Japan Research Institute, Ltd., Tokyo (Japan)

    2001-06-01

    It is well known that a composition of spent fuel depends on various parameter changes throughout a burn-up period. In this study we aimed at the boron concentration and its change, the coolant temperature and its spatial distribution, the specific power, the operation mode, and the duration of inspection, because the effects due to these parameters have not been analyzed in detail. The composition changes of spent fuel were calculated by using the burn-up code SWAT, when the parameters mentioned above varied in the range of actual variations. Moreover, to estimate the reactivity effect caused by the composition changes, the criticality calculations for an infinite array of spent fuel were carried out with computer codes SRAC95 or MVP. In this report the reactivity effects were arranged from the viewpoint of what parameters gave more positive reactivity effect. The results obtained through this study are useful to choose the burn-up calculation model when we take account of the burn-up credit in the spent fuel management. (author)

  11. Depletion of gadolinium burnable poison in a PWR assembly with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Refeat, Riham Mahmoud [Nuclear and Radiological Regulatory Authority (NRRA), Cairo (Egypt). Safety Engineering Dept.

    2015-12-15

    A tendency to increase the discharge burnup of nuclear fuel for Advanced Pressurized Water Reactors (PWR) has been a characteristic of its operation for many years. It will be able to burn at very high burnup of about 70 GWd/t with UO{sub 2} fuels. The U-235 enrichment must be higher than 5 %, which leads to the necessity of using an extremely efficient burnable poison like Gadolinium oxide. Using gadolinium isotope is significant due to its particular depletion behavior (''Onion-Skin'' effect). In this paper, the MCNPX2.7 code is used to calculate the important neutronic parameters of the next generation fuels of PWR. K-infinity, local peaking factor and fission rate distributions are calculated for a PWR assembly which burn at very high burnup reaching 70 GWd/t. The calculations are performed using the recently released evaluated Gadolinium cross section data. The results obtained are close to those of a LWR next generation fuel benchmark problem. This demonstrates that the calculation scheme used is able to accurately model a PWR assembly that operates at high burnup values.

  12. Analysis of bubble pressure in the rim region of high burnup PWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Yang Hyun; Lee, Byung Ho; Sohn, Dong Seong [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-02-01

    Bubble pressure in the rim region of high burnup PWR UO{sub 2} fuel has been modeled based on measured rim width, porosity and bubble density. Using the assumption that excessive bubble pressure in the rim is inversely proportional to its radius, proportionality constant is derived as a function of average pellet burnup and bubble radius. This approach is possible because the integration of the number of Xe atoms retained in the rim bubbles, which can be calculated as a function of bubble radius, over the bubble radius gives the total number of Xe atoms in the rim bubbles. Here the total number of Xe atoms in the rim bubbles can be derived from the measured Xe depletion fraction in the matrix and the calculated rim thickness. Then the rim bubble pressure is obtained as a function of fuel burnup and bubble size from the proportionality constant. Therefore, the present model can provide some useful information that would be required to analyze the behavior of high burnup PWR UO{sub 2} fuel under both normal and transient operating conditions. 28 refs., 9 figs. (Author)

  13. Thermal properties of U–Mo alloys irradiated to moderate burnup and power

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas E., E-mail: Douglas.Burkes@pnnl.gov; Casella, Andrew M.; Casella, Amanda J.; Buck, Edgar C.; Pool, Karl N.; MacFarlan, Paul J.; Edwards, Matthew K.; Smith, Frances N.

    2015-09-15

    Highlights: • Thermal properties of irradiated U–Mo alloy monolithic fuel samples were measured. • Density, thermal diffusivity, and thermal conductivity are influenced by increasing burnup. • U–Mo chemistry and specific heat capacity was not as sensitive to increasing burnup. • Thermal conductivity decreased approximately 45% for a fission density of 4.52 × 10{sup 21} fissions cm{sup −3} at 200 °C. • An empirical model developed previously agrees well with the experimental measurements. - Abstract: A variety of physical and thermal property measurements as a function of temperature and fission density were performed on irradiated U–Mo alloy monolithic fuel samples with a Zr diffusion barrier and clad in aluminum alloy 6061. The U–Mo alloy density, thermal diffusivity, and thermal conductivity are strongly influenced by increasing burnup, mainly as the result of irradiation induced recrystallization and fission gas bubble formation and coalescence. U–Mo chemistry, specifically Mo content, and specific heat capacity was not as sensitive to increasing burnup. Measurements indicated that thermal conductivity of the U–Mo alloy decreased approximately 30% for a fission density of 3.30 × 10{sup 21} fissions cm{sup −3} and approximately 45% for a fission density of 4.52 × 10{sup 21} fissions cm{sup −3} from unirradiated values at 200 °C. An empirical thermal conductivity degradation model developed previously and summarized here agrees well with the experimental measurements.

  14. Oxygen potential measurements in high burnup LWR U0 2 fuel

    Science.gov (United States)

    Matzke, Hj.

    1995-05-01

    A miniature solid state galvanic cell was used to measure the oxygen potential Δ overlineG( O2) of reactor irradiated U0 2 fuel at different burnups in the range of 28 to ⩾ 150 GWd d/t M. This very high burnup was achieved in the rim region of a fuel with a cross section average burnup of 75 GWd d/t M. The fuels had different enrichments and therefore different contributions of fission of 235U and 239Pu. The temperature range covered was 900 to 1350 K. None of the fuels showed a significant oxidation. Rather, if allowance is made for the dissolved rare earth fission products and the Pu formed during irradiation, some of the fuels were very slightly substoichiometric and the highest possible degree of oxidation corresponded to U0 2.001. In general, the Δ overlineG( O2) at 750°C was about -400 kJ/mol, corresponding to the Δ overlineG( O2) of the reaction Mo + O 2 → MoO 2. The implication of these results which are in contrast to commonly assumed ideas that U0 2 fuel oxidizes due to burnup, are discussed and the importance of the fission product Mo and of the zircaloy clad as oxygen buffers is outlined.

  15. Radiochemical Assays of Irradiated VVER-440 Fuel for Use in Spent Fuel Burnup Credit Activities

    Energy Technology Data Exchange (ETDEWEB)

    Jardine, L J

    2005-04-25

    The objective of this spent fuel burnup credit work was to study and describe a VVER-440 reactor spent fuel assembly (FA) initial state before irradiation, its operational irradiation history and the resulting radionuclide distribution in the fuel assembly after irradiation. This work includes the following stages: (1) to pick out and select a specific spent (irradiated) FA for examination; (2) to describe the FA initial state before irradiation; (3) to describe the irradiation history, including thermal calculations; (4) to examine the burnup distribution of select radionuclides along the FA height and cross-section; (5) to examine the radionuclide distributions; (6) to determine the Kr-85 release into the plenum; (7) to select and prepare FA rod specimens for destructive examinations; (8) to determine the radionuclide compositions, isotope masses and burnup in the rod specimens; and (9) to analyze, document and process the results. The specific workscope included the destructive assay (DA) of spent fuel assembly rod segments with an {approx}38.5 MWd/KgU burnup from a single VVER-440 fuel assembly from the Novovorenezh reactor in Russia. Based on irradiation history criteria, four rods from the fuel assembly were selected and removed from the assembly for examination. Next, 8 sections were cut from the four rods and sent for destructive analysis of radionuclides by radiochemical analyses. The results were documented in a series of seven reports over a period of {approx}1 1/2 years.

  16. Parametric Study of the Effect of Burnable Poison Rods for PWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J.C.

    2001-09-28

    The Interim Staff Guidance on burnup credit (ISG-8) issued by the United States Nuclear Regulatory Commission's (U.S. NRC) Spent Fuel Project Office recommends restricting the use of burnup credit to assemblies that have not used burnable absorbers. This recommended restriction eliminates a large portion of the currently discharged spent fuel assemblies from cask loading, and thus severely limits the practical usefulness of burnup credit. In the absence of readily available information on burnable poison rod (BPR) design specifications and usage in U.S. pressurized-water-reactors (PWRs), and the subsequent reactivity effect of BPR exposure on discharged spent nuclear fuel (SNF), NRC staff has indicated a need for additional information in these areas. In response, this report presents a parametric study of the effect of BPR exposure on the reactivity of SNF for various BPR designs, fuel enrichments, and exposure conditions, and documents BPR design specifications. Trends in the reactivity effects of BPRs are established with infinite pin-cell and assembly array calculations with the SCALE and HELIOS code packages, respectively. Subsequently, the reactivity effects of BPRs for typical initial enrichment and burnup combinations are quantified based on three-dimensional (3-D) KENO V.a Monte Carlo calculations with a realistic rail-type cask designed for burnup credit. The calculations demonstrate that the positive reactivity effect due to BPR exposure increases nearly linearly with burnup and is dependent on the number, poison loading, and design of the BPRs and the initial fuel enrichment. Expected typical reactivity increases, based on one-cycle BPR exposure, were found to be less than 1% {Delta}k. Based on the presented analysis, guidance is offered on an appropriate approach for calculating bounding SNF isotopic data for assemblies exposed to BPRs. Although the analyses do not address the issue of validation of depletion methods for assembly designs with BPRs

  17. Functionalized boron nitride nanotubes

    Science.gov (United States)

    Sainsbury, Toby; Ikuno, Takashi; Zettl, Alexander K

    2014-04-22

    A plasma treatment has been used to modify the surface of BNNTs. In one example, the surface of the BNNT has been modified using ammonia plasma to include amine functional groups. Amine functionalization allows BNNTs to be soluble in chloroform, which had not been possible previously. Further functionalization of amine-functionalized BNNTs with thiol-terminated organic molecules has also been demonstrated. Gold nanoparticles have been self-assembled at the surface of both amine- and thiol-functionalized boron nitride Nanotubes (BNNTs) in solution. This approach constitutes a basis for the preparation of highly functionalized BNNTs and for their utilization as nanoscale templates for assembly and integration with other nanoscale materials.

  18. Propagation of statistical and nuclear data uncertainties in Monte Carlo burn-up calculations

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Herranz, Nuria [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid, UPM (Spain)], E-mail: nuria@din.upm.es; Cabellos, Oscar [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid, UPM (Spain); Sanz, Javier [Departamento de Ingenieria Energetica, Universidad Nacional de Educacion a Distancia, UNED (Spain); Juan, Jesus [Laboratorio de Estadistica, Universidad Politecnica de Madrid, UPM (Spain); Kuijper, Jim C. [NRG - Fuels, Actinides and Isotopes Group, Petten (Netherlands)

    2008-04-15

    Two methodologies to propagate the uncertainties on the nuclide inventory in combined Monte Carlo-spectrum and burn-up calculations are presented, based on sensitivity/uncertainty and random sampling techniques (uncertainty Monte Carlo method). Both enable the assessment of the impact of uncertainties in the nuclear data as well as uncertainties due to the statistical nature of the Monte Carlo neutron transport calculation. The methodologies are implemented in our MCNP-ACAB system, which combines the neutron transport code MCNP-4C and the inventory code ACAB. A high burn-up benchmark problem is used to test the MCNP-ACAB performance in inventory predictions, with no uncertainties. A good agreement is found with the results of other participants. This benchmark problem is also used to assess the impact of nuclear data uncertainties and statistical flux errors in high burn-up applications. A detailed calculation is performed to evaluate the effect of cross-section uncertainties in the inventory prediction, taking into account the temporal evolution of the neutron flux level and spectrum. Very large uncertainties are found at the unusually high burn-up of this exercise (800 MWd/kgHM). To compare the impact of the statistical errors in the calculated flux with respect to the cross uncertainties, a simplified problem is considered, taking a constant neutron flux level and spectrum. It is shown that, provided that the flux statistical deviations in the Monte Carlo transport calculation do not exceed a given value, the effect of the flux errors in the calculated isotopic inventory are negligible (even at very high burn-up) compared to the effect of the large cross-section uncertainties available at present in the data files.

  19. OECD/NEA Burnup Credit Calculational Criticality Benchmark Phase I-B Results

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    1993-01-01

    Burnup credit is an ongoing technical concern for many countries that operate commercial nuclear power reactors. In a multinational cooperative effort to resolve burnup credit issues, a Burnup Credit Working Group has been formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development. This working group has established a set of well-defined calculational benchmarks designed to study significant aspects of burnup credit computational methods. These benchmarks are intended to provide a means for the intercomparison of computer codes, methods, and data applied in spent fuel analysis. The benchmarks have been divided into multiple phases, each phase focusing on a particular feature of burnup credit analysis. This report summarizes the results and findings of the Phase I-B benchmark, which was proposed to provide a comparison of the ability of different code systems and data libraries to perform depletion analysis for the prediction of spent fuel isotopic concentrations. Results included here represent 21 different sets of calculations submitted by 16 different organizations worldwide, and are based on a limited set of nuclides determined to have the most important effect on the neutron multiplication factor of light-water-reactor spent fuel. A comparison of all sets of results demonstrates that most methods are in agreement to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods are within 11% agreement about the average for all fission products studied. Furthermore, most deviations are less than 10%, and many are less than 5%. The exceptions are {sup 149}Sm, {sup 151}Sm, and {sup 155}Gd.

  20. OECD/NEA burnup credit calculational criticality benchmark Phase I-B results

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.; Parks, C.V. [Oak Ridge National Lab., TN (United States); Brady, M.C. [Sandia National Labs., Las Vegas, NV (United States)

    1996-06-01

    In most countries, criticality analysis of LWR fuel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. This assumption has led to the design of widely spaced and/or highly poisoned storage and transport arrays. If credit is assumed for fuel burnup, initial enrichment limitations can be raised in existing systems, and more compact and economical arrays can be designed. Such reliance on the reduced reactivity of spent fuel for criticality control is referred to as burnup credit. The Burnup Credit Working Group, formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development, has established a set of well-defined calculational benchmarks designed to study significant aspects of burnup credit computational methods. These benchmarks are intended to provide a means for the intercomparison of computer codes, methods, and data applied in spent fuel analysis. The benchmarks have been divided into multiple phases, each phase focusing on a particular feature of burnup credit analysis. This report summarizes the results and findings of the Phase I-B benchmark, which was proposed to provide a comparison of the ability of different code systems and data libraries to perform depletion analysis for the prediction of spent fuel isotopic concentrations. Results included here represent 21 different sets of calculations submitted by 16 different organizations worldwide and are based on a limited set of nuclides determined to have the most important effect on the neutron multiplication factor of light-water-reactor spent fuel. A comparison of all sets of results demonstrates that most methods agree to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods agree within 11% about the average for all fission products studied. Most deviations are less than 10%, and many are less than 5%. The exceptions are Sm 149, Sm 151, and Gd 155.

  1. Application of Genetic Algorithm methodologies in fuel bundle burnup optimization of Pressurized Heavy Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jayalal, M.L., E-mail: jayalal@igcar.gov.in [Electronics, Instrumentation and Radiological Safety Group (EIRSG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India); Ramachandran, Suja [Electronics, Instrumentation and Radiological Safety Group (EIRSG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India); Rathakrishnan, S. [Reactor Physics Section, Madras Atomic Power Station (MAPS), Kalpakkam, Tamil Nadu (India); Satya Murty, S.A.V. [Electronics, Instrumentation and Radiological Safety Group (EIRSG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India); Sai Baba, M. [Resources Management Group (RMG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India)

    2015-01-15

    Highlights: • We study and compare Genetic Algorithms (GA) in the fuel bundle burnup optimization of an Indian Pressurized Heavy Water Reactor (PHWR) of 220 MWe. • Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are considered. • For the selected problem, Multi Objective GA performs better than Penalty Functions based GA. • In the present study, Multi Objective GA outperforms Penalty Functions based GA in convergence speed and better diversity in solutions. - Abstract: The work carried out as a part of application and comparison of GA techniques in nuclear reactor environment is presented in the study. The nuclear fuel management optimization problem selected for the study aims at arriving appropriate reference discharge burnup values for the two burnup zones of 220 MWe Pressurized Heavy Water Reactor (PHWR) core. Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are applied in this study. The study reveals, for the selected problem of PHWR fuel bundle burnup optimization, Multi Objective GA is more suitable than Penalty Functions based GA in the two aspects considered: by way of producing diverse feasible solutions and the convergence speed being better, i.e. it is capable of generating more number of feasible solutions, from earlier generations. It is observed that for the selected problem, the Multi Objective GA is 25.0% faster than Penalty Functions based GA with respect to CPU time, for generating 80% of the population with feasible solutions. When average computational time of fixed generations are considered, Penalty Functions based GA is 44.5% faster than Multi Objective GA. In the overall performance, the convergence speed of Multi Objective GA surpasses the computational time advantage of Penalty Functions based GA. The ability of Multi Objective GA in producing more diverse feasible solutions is a desired feature of the problem selected, that helps the

  2. High Burnup Dry Storage Cask Research and Development Project, Final Test Plan

    Energy Technology Data Exchange (ETDEWEB)

    None

    2014-02-27

    EPRI is leading a project team to develop and implement the first five years of a Test Plan to collect data from a SNF dry storage system containing high burnup fuel.12 The Test Plan defined in this document outlines the data to be collected, and the storage system design, procedures, and licensing necessary to implement the Test Plan.13 The main goals of the proposed test are to provide confirmatory data14 for models, future SNF dry storage cask design, and to support license renewals and new licenses for ISFSIs. To provide data that is most relevant to high burnup fuel in dry storage, the design of the test storage system must mimic real conditions that high burnup SNF experiences during all stages of dry storage: loading, cask drying, inert gas backfilling, and transfer to the ISFSI for multi-year storage.15 Along with other optional modeling, SETs, and SSTs, the data collected in this Test Plan can be used to evaluate the integrity of dry storage systems and the high burnup fuel contained therein over many decades. It should be noted that the Test Plan described in this document discusses essential activities that go beyond the first five years of Test Plan implementation.16 The first five years of the Test Plan include activities up through loading the cask, initiating the data collection, and beginning the long-term storage period at the ISFSI. The Test Plan encompasses the overall project that includes activities that may not be completed until 15 or more years from now, including continued data collection, shipment of the Research Project Cask to a Fuel Examination Facility, opening the cask at the Fuel Examination Facility, and examining the high burnup fuel after the initial storage period.

  3. Effects of Aqueous Vapour Consistence in Nitriding Furnace on the Quality of the Sintered Nitride

    Institute of Scientific and Technical Information of China (English)

    WANGZijiang

    1998-01-01

    If the aqueous vapour consistence is too high(>0.7%),it is very disadvantageous to the sintered products in the nitriding furnace,when silcon nitride bonded silicon carbide products are synthesized by nitridation of silicon.

  4. Behaviour of fission gas in the rim region of high burn-up UO 2 fuel pellets with particular reference to results from an XRF investigation

    Science.gov (United States)

    Mogensen, M.; Pearce, J. H.; Walker, C. T.

    1999-01-01

    XRF and EPMA results for retained xenon from Battelle's high burn-up effects program are re-evaluated. The data reviewed are from commercial low enriched BWR fuel with burn-ups of 44.8-54.9 GWd/tU and high enriched PWR fuel with burn-ups from 62.5 to 83.1 GWd/tU. It is found that the high burn-up structure penetrated much deeper than initially reported. The local burn-up threshold for the formation of the high burn-up structure in those fuels with grain sizes in the normal range lay between 60 and 75 GWd/tU. The high burn-up structure was not detected by EPMA in a fuel that had a grain size of 78 μm although the local burn-up at the pellet rim had exceeded 80 GWd/tU. It is concluded that fission gas had been released from the high burn-up structure in three PWR fuel sections with burn-ups of 70.4, 72.2 and 83.1 GWd/tU. In the rim region of the last two sections at the locations where XRF indicated gas release the local burn-up was higher than 75 GWd/tU.

  5. Inter-layer potential for hexagonal boron nitride

    Energy Technology Data Exchange (ETDEWEB)

    Leven, Itai; Hod, Oded, E-mail: odedhod@tau.ac.il [Department of Chemical Physics, School of Chemistry, The Raymond and Beverly Sackler Faculty of Exact Sciences, Tel-Aviv University, Tel-Aviv 69978 (Israel); Azuri, Ido; Kronik, Leeor [Department of Materials and Interfaces, Weizmann Institute of Science, Rehovoth 76100 (Israel)

    2014-03-14

    A new interlayer force-field for layered hexagonal boron nitride (h-BN) based structures is presented. The force-field contains three terms representing the interlayer attraction due to dispersive interactions, repulsion due to anisotropic overlaps of electron clouds, and monopolar electrostatic interactions. With appropriate parameterization, the potential is able to simultaneously capture well the binding and lateral sliding energies of planar h-BN based dimer systems as well as the interlayer telescoping and rotation of double walled boron-nitride nanotubes of different crystallographic orientations. The new potential thus allows for the accurate and efficient modeling and simulation of large-scale h-BN based layered structures.

  6. Communication: Water on hexagonal boron nitride from diffusion Monte Carlo

    Energy Technology Data Exchange (ETDEWEB)

    Al-Hamdani, Yasmine S.; Ma, Ming; Michaelides, Angelos, E-mail: angelos.michaelides@ucl.ac.uk [Thomas Young Centre and London Centre for Nanotechnology, 17–19 Gordon Street, London WC1H 0AH (United Kingdom); Department of Chemistry, University College London, 20 Gordon Street, London WC1H 0AJ (United Kingdom); Alfè, Dario [Thomas Young Centre and London Centre for Nanotechnology, 17–19 Gordon Street, London WC1H 0AH (United Kingdom); Department of Earth Sciences, University College London, Gower Street, London WC1E 6BT (United Kingdom); Lilienfeld, O. Anatole von [Institute of Physical Chemistry and National Center for Computational Design and Discovery of Novel Materials, Department of Chemistry, University of Basel, Klingelbergstrasse 80, CH-4056 Basel (Switzerland); Argonne Leadership Computing Facility, Argonne National Laboratories, 9700 S. Cass Avenue Argonne, Lemont, Illinois 60439 (United States)

    2015-05-14

    Despite a recent flurry of experimental and simulation studies, an accurate estimate of the interaction strength of water molecules with hexagonal boron nitride is lacking. Here, we report quantum Monte Carlo results for the adsorption of a water monomer on a periodic hexagonal boron nitride sheet, which yield a water monomer interaction energy of −84 ± 5 meV. We use the results to evaluate the performance of several widely used density functional theory (DFT) exchange correlation functionals and find that they all deviate substantially. Differences in interaction energies between different adsorption sites are however better reproduced by DFT.

  7. Inter-layer potential for hexagonal boron nitride

    Science.gov (United States)

    Leven, Itai; Azuri, Ido; Kronik, Leeor; Hod, Oded

    2014-03-01

    A new interlayer force-field for layered hexagonal boron nitride (h-BN) based structures is presented. The force-field contains three terms representing the interlayer attraction due to dispersive interactions, repulsion due to anisotropic overlaps of electron clouds, and monopolar electrostatic interactions. With appropriate parameterization, the potential is able to simultaneously capture well the binding and lateral sliding energies of planar h-BN based dimer systems as well as the interlayer telescoping and rotation of double walled boron-nitride nanotubes of different crystallographic orientations. The new potential thus allows for the accurate and efficient modeling and simulation of large-scale h-BN based layered structures.

  8. Burnup calculation by the method of first-flight collision probabilities using average chords prior to the first collision

    Science.gov (United States)

    Karpushkin, T. Yu.

    2012-12-01

    A technique to calculate the burnup of materials of cells and fuel assemblies using the matrices of first-flight neutron collision probabilities rebuilt at a given burnup step is presented. A method to rebuild and correct first collision probability matrices using average chords prior to the first neutron collision, which are calculated with the help of geometric modules of constructed stochastic neutron trajectories, is described. Results of calculation of the infinite multiplication factor for elementary cells with a modified material composition compared to the reference one as well as calculation of material burnup in the cells and fuel assemblies of a VVER-1000 are presented.

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

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, William BJ J [ORNL; Ade, Brian J [ORNL; Bowman, Stephen M [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Mertyurek, Ugur [ORNL; Radulescu, Georgeta [ORNL

    2015-01-01

    Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (keff) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of lattice design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup

  10. Study of the Active Screen Plasma Nitriding

    Institute of Scientific and Technical Information of China (English)

    Zhao Cheng; C. X. Li; H. Dong; T. Bell

    2004-01-01

    Active screen plasma nitriding (ASPN) is a novel nitriding process, which overcomes many of the practical problems associated with the conventional DC plasma nitriding (DCPN). Experimental results showed that the metallurgical characteristics and hardening effect of 722M24 steel nitrided by ASPN at both floating potential and anodic (zero) potential were similar to those nitrided by DCPN. XRD and high-resolution SEM analysis indicated that iron nitride particles with sizes in sub-micron scale were deposited on the specimen surface in AS plasma nitriding. These indicate that the neutral iron nitride particles, which are sputtered from the active screen and transferred through plasma to specimen surface, are considered to be the dominant nitrogen carder in ASPN. The OES results show that NH could not be a critical species in plasma nitriding.

  11. Development of a Burnup Module DECBURN Based on the Krylov Subspace Method

    Energy Technology Data Exchange (ETDEWEB)

    Cho, J. Y.; Kim, K. S.; Shim, H. J.; Song, J. S

    2008-05-15

    This report is to develop a burnup module DECBURN that is essential for the reactor analysis and the assembly homogenization codes to trace the fuel composition change during the core burnup. The developed burnup module solves the burnup equation by the matrix exponential method based on the Krylov Subspace method. The final solution of the matrix exponential is obtained by the matrix scaling and squaring method. To develop DECBURN module, this report includes the followings as: (1) Krylov Subspace Method for Burnup Equation, (2) Manufacturing of the DECBURN module, (3) Library Structure Setup and Library Manufacturing, (4) Examination of the DECBURN module, (5) Implementation to the DeCART code and Verification. DECBURN library includes the decay constants, one-group cross section and the fission yields. Examination of the DECBURN module is performed by manufacturing a driver program, and the results of the DECBURN module is compared with those of the ORIGEN program. Also, the implemented DECBURN module to the DeCART code is applied to the LWR depletion benchmark and a OPR-1000 pin cell problem, and the solutions are compared with the HELIOS code to verify the computational soundness and accuracy. In this process, the criticality calculation method and the predictor-corrector scheme are introduced to the DeCART code for a function of the homogenization code. The examination by a driver program shows that the DECBURN module produces exactly the same solution with the ORIGEN program. DeCART code that equips the DECBURN module produces a compatible solution to the other codes for the LWR depletion benchmark. Also the multiplication factors of the DeCART code for the OPR-1000 pin cell problem agree to the HELIOS code within 100 pcm over the whole burnup steps. The multiplication factors with the criticality calculation are also compatible with the HELIOS code. These results mean that the developed DECBURN module works soundly and produces an accurate solution

  12. Synthesis of ternary metal nitride nanoparticles using mesoporous carbon nitride as reactive template.

    Science.gov (United States)

    Fischer, Anna; Müller, Jens Oliver; Antonietti, Markus; Thomas, Arne

    2008-12-23

    Mesoporous graphitic carbon nitride was used as both a nanoreactor and a reactant for the synthesis of ternary metal nitride nanoparticles. By infiltration of a mixture of two metal precursors into mesoporous carbon nitride, the pores act first as a nanoconfinement, generating amorphous mixed oxide nanoparticles. During heating and decomposition, the carbon nitride second acts as reactant or, more precisely, as a nitrogen source, which converts the preformed mixed oxide nanoparticles into the corresponding nitride (reactive templating). Using this approach, ternary metal nitride particles with diameters smaller 10 nm composed of aluminum gallium nitride (Al-Ga-N) and titanium vanadium nitride (Ti-V-N) were synthesized. Due to the confinement effect of the carbon nitride matrix, the composition of the resulting metal nitride can be easily adjusted by changing the concentration of the preceding precursor solution. Thus, ternary metal nitride nanoparticles with continuously adjustable metal composition can be produced.

  13. Boron nitride converted carbon fiber

    Science.gov (United States)

    Rousseas, Michael; Mickelson, William; Zettl, Alexander K.

    2016-04-05

    This disclosure provides systems, methods, and apparatus related to boron nitride converted carbon fiber. In one aspect, a method may include the operations of providing boron oxide and carbon fiber, heating the boron oxide to melt the boron oxide and heating the carbon fiber, mixing a nitrogen-containing gas with boron oxide vapor from molten boron oxide, and converting at least a portion of the carbon fiber to boron nitride.

  14. Economic incentives and recommended development for commercial use of high burnup fuels in the once-through LWR fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Stout, R.B.; Merckx, K.R.; Holm, J.S.

    1981-01-01

    This study calculates the reduced uranium requirements and the economic incentives for increasing the burnup of current design LWR fuels from the current range of 25 to 35 MWD/Kg to a range of 45 to 55 MWD/Kg. The changes in fuel management strategies which may be required to accommodate these high burnup fuels and longer fuel cycles are discussed. The material behavior problems which may present obstacles to achieving high burnup or to license fuel are identified and discussed. These problems are presented in terms of integral fuel response and the informational needs for commercial and licensing acceptance. Research and development programs are outlined which are aimed at achieving a licensing position and commercial acceptance of high burnup fuels.

  15. The Challenges Associated with High Burnup and High Temperature for UO2 TRISO-Coated Particle Fuel

    Energy Technology Data Exchange (ETDEWEB)

    David Petti; John Maki

    2005-02-01

    The fuel service conditions for the DOE Next Generation Nuclear Plant (NGNP) will be challenging. All major fuel related design parameters (burnup, temperature, fast neutron fluence, power density, particle packing fraction) exceed the values that were qualified in the successful German UO2 TRISO-coated particle fuel development program in the 1980s. While TRISO-coated particle fuel has been irradiated at NGNP relevant levels for two or three of the design parameters, no data exist for TRISO-coated particle fuel for all five parameters simultaneously. Of particular concern are the high burnup and high temperatures expected in the NGNP. In this paper, where possible, we evaluate the challenges associated with high burnup and high temperature quantitatively by examining the performance of the fuel in terms of different known failure mechanisms. Potential design solutions to ameliorate the negative effects of high burnup and high temperature are also discussed.

  16. Modeling of PWR fuel at extended burnup; Estudo de modelos para o comportamento a altas queimas de varetas combustiveis de reatores a agua leve pressurizada

    Energy Technology Data Exchange (ETDEWEB)

    Dias, Raphael Mejias

    2016-11-01

    This work studies the modifications implemented over successive versions in the empirical models of the computer program FRAPCON used to simulate the steady state irradiation performance of Pressurized Water Reactor (PWR) fuel rods under high burnup condition. In the study, the empirical models present in FRAPCON official documentation were analyzed. A literature study was conducted on the effects of high burnup in nuclear fuels and to improve the understanding of the models used by FRAPCON program in these conditions. A steady state fuel performance analysis was conducted for a typical PWR fuel rod using FRAPCON program versions 3.3, 3.4, and 3.5. The results presented by the different versions of the program were compared in order to verify the impact of model changes in the output parameters of the program. It was observed that the changes brought significant differences in the results of the fuel rod thermal and mechanical parameters, especially when they evolved from FRAPCON-3.3 version to FRAPCON-3.5 version. Lower temperatures, lower cladding stress and strain, lower cladding oxide layer thickness were obtained in the fuel rod analyzed with the FRAPCON-3.5 version. (author)

  17. Hemocompatibility of titanium nitride.

    Science.gov (United States)

    Dion, I; Baquey, C; Candelon, B; Monties, J R

    1992-10-01

    The left ventricular assist device is based on the principle of the Maillard-Wenkel rotative pump. The materials which make up the pump must present particular mechanical, tribological, thermal and chemical properties. Titanium nitride (TiN) because of its surface properties and graphite because of its bulk characteristics have been chosen. The present study evaluated the in vitro hemocompatibility of TiN coating deposited by the chemical vapor deposition process. Protein adsorption, platelet retention and hemolysis tests have been carried out. In spite of some disparities, the TiN behavior towards albumin and fibrinogen is interesting, compared with the one of a reference medical grade elastomer. The platelet retention test gives similar results as those achieved with the same elastomer. The hemolysis percentage is near to zero. TiN shows interesting characteristics, as far as mechanical and tribological problems are concerned, and presents very encouraging blood tolerability properties.

  18. Hardening Roll Surface by Plasma Nitriding with Subsequent Hardfacing

    Science.gov (United States)

    Pesin, A.; Pustovoytov, D.; Vafin, R.; Yagafarov, I.; Vardanyan, E.

    2017-05-01

    The wear of the surface layer of rolls after ion nitriding in glow discharge, followed by a coating of TiN -TiAlN plasma arc are studied and simulated. stress-strain state of the material rolls under asymmetric rolling with ultra-high shear deformations is simulated. The effect of thermal fields, formed upon contact of the tool and a deformable sheet, the structure of aluminum alloys, are considered.

  19. FRAPCON-3: Modifications to fuel rod material properties and performance models for high-burnup application

    Energy Technology Data Exchange (ETDEWEB)

    Lanning, D.D.; Beyer, C.E.; Painter, C.L.

    1997-12-01

    This volume describes the fuel rod material and performance models that were updated for the FRAPCON-3 steady-state fuel rod performance code. The property and performance models were changed to account for behavior at extended burnup levels up to 65 Gwd/MTU. The property and performance models updated were the fission gas release, fuel thermal conductivity, fuel swelling, fuel relocation, radial power distribution, solid-solid contact gap conductance, cladding corrosion and hydriding, cladding mechanical properties, and cladding axial growth. Each updated property and model was compared to well characterized data up to high burnup levels. The installation of these properties and models in the FRAPCON-3 code along with input instructions are provided in Volume 2 of this report and Volume 3 provides a code assessment based on comparison to integral performance data. The updated FRAPCON-3 code is intended to replace the earlier codes FRAPCON-2 and GAPCON-THERMAL-2. 94 refs., 61 figs., 9 tabs.

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

    Science.gov (United States)

    Lemoine, F.

    1997-09-01

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

  1. Draft evaluation of the frequency for gas sampling for the high burnup confirmatory data project

    Energy Technology Data Exchange (ETDEWEB)

    Stockman, Christine T. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Alsaed, Halim A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-03-26

    This report fulfills the M3 milestone M3FT-15SN0802041, “Draft Evaluation of the Frequency for Gas Sampling for the High Burn-up Storage Demonstration Project” under Work Package FT-15SN080204, “ST Field Demonstration Support – SNL”. This report provides a technically based gas sampling frequency strategy for the High Burnup (HBU) Confirmatory Data Project. The evaluation of: 1) the types and magnitudes of gases that could be present in the project cask and, 2) the degradation mechanisms that could change gas compositions culminates in an adaptive gas sampling frequency strategy. This adaptive strategy is compared against the sampling frequency that has been developed based on operational considerations. Gas sampling will provide information on the presence of residual water (and byproducts associated with its reactions and decomposition) and breach of cladding, which could inform the decision of when to open the project cask.

  2. Experience with incomplete control rod insertion in fuel with burnup exceeding approximately 40 GWD/MTU

    Energy Technology Data Exchange (ETDEWEB)

    Kee, E. [Houston Lighting & Power Co., Wadworth, TX (United States)

    1997-01-01

    Analysis and measurement experience with fuel assemblies having incomplete control rod insertion at burnups of approximately 40 GWD/MTU is presented. Control rod motion dynamics and simplified structural analyses are presented and compared to measurement data. Fuel assembly growth measurements taken with the plant Refueling Machine Z-Tape are described and presented. Bow measurements (including plug gauging) are described and potential improvements are suggested. The measurements described and analysis performed show that sufficient guide tube bow (either from creep or yield buckling) is present in some high burnup assemblies to stop the control rods before they reach their full limit of travel. Recommendations are made that, if implemented, could improve cost performance related to testing and analysis activities.

  3. Determination of deuterium–tritium critical burn-up parameter by four temperature theory

    Energy Technology Data Exchange (ETDEWEB)

    Nazirzadeh, M.; Ghasemizad, A. [Department of Physics, University of Guilan, 41335-1914 Rasht (Iran, Islamic Republic of); Khanbabei, B. [School of Physics, Damghan University, 36716-41167 Damghan (Iran, Islamic Republic of)

    2015-12-15

    Conditions for thermonuclear burn-up of an equimolar mixture of deuterium-tritium in non-equilibrium plasma have been investigated by four temperature theory. The photon distribution shape significantly affects the nature of thermonuclear burn. In three temperature model, the photon distribution is Planckian but in four temperature theory the photon distribution has a pure Planck form below a certain cut-off energy and then for photon energy above this cut-off energy makes a transition to Bose-Einstein distribution with a finite chemical potential. The objective was to develop four temperature theory in a plasma to calculate the critical burn up parameter which depends upon initial density, the plasma components initial temperatures, and hot spot size. All the obtained results from four temperature theory model are compared with 3 temperature model. It is shown that the values of critical burn-up parameter calculated by four temperature theory are smaller than those of three temperature model.

  4. The Impact of Operating Parameters and Correlated Parameters for Extended BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Marshall, William B. J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ilas, Germina [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bowman, Stephen M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-06-01

    Applicants for certificates of compliance for spent nuclear fuel (SNF) transportation and dry storage systems perform analyses to demonstrate that these systems are adequately subcritical per the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Parts 71 and 72. For pressurized water reactor (PWR) SNF, these analyses may credit the reduction in assembly reactivity caused by depletion of fissile nuclides and buildup of neutron-absorbing nuclides during power operation. This credit for reactivity reduction during depletion is commonly referred to as burnup credit (BUC). US Nuclear Regulatory Commission (NRC) staff review BUC analyses according to the guidance in the Division of Spent Fuel Storage and Transportation Interim Staff Guidance (ISG) 8, Revision 3, Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transportation and Storage Casks.

  5. Thermal properties of U–Mo alloys irradiated to moderate burnup and power

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-01

    A variety of physical and thermal property measurements as a function of temperature and fission density were performed on irradiated U-Mo alloy monolithic fuel samples with a Zr diffusion barrier and clad in aluminum alloy 6061. The U-Mo alloy density, thermal diffusivity, and thermal conductivity are strongly influenced by increasing burnup, mainly as the result of irradiation induced recrystallization and fission gas bubble formation and coalescence. U-Mo chemistry, specifically Mo content, and specific heat capacity was not as sensitive to increasing burnup. Measurements indicated that thermal conductivity of the U-Mo alloy decreased approximately 30% for a fission density of 2.88 × 1021 fissions cm-3 and approximately 45% for a fission density of 4.08 × 1021 fissions cm-3 from unirradiated values at 200 oC. An empirical thermal conductivity degradation model developed previously and summarized here agrees well with the experimental measurements.

  6. Assessing the Effect of Fuel Burnup on Control Rod Worth for HEU and LEU Cores of Gharr-1

    Directory of Open Access Journals (Sweden)

    E.K. Boafo

    2013-02-01

    Full Text Available An important parameter in the design and analysis of a nuclear reactor is the reactivity worth of the control rod which is a measure of the efficiency of the control rod to absorb excess reactivity. During reactor operation, the control rod worth is affected by factors such as the fuel burnup, Xenon concentration, Samarium concentration and the position of the control rod in the core. This study investigates the effect of fuel burnup on the control rod worth by comparing results of a fresh and an irradiated core of Ghana's Miniature Neutron Source Reactor for both HEU and LEU cores. In this study, two codes have been utilized namely BURNPRO for fuel burnup calculation and MCNP5 which uses densities of actinides of the irradiated fuel obtained from BURNPRO. Results showed a decrease of the control rod worth with burnup for the LEU while rod worth increased with burnup for the HEU core. The average thermal flux in both inner and outer irradiation sites also decreased significantly with burnup for both cores.

  7. Evaluation of safety criteria on LOCA and RIA for high burnup nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sun Ki; Bang, Je Geon; Kim, Dae Ho; Yang, Yong Sik; Song, Keun Woo

    2007-12-15

    Comprehensive researches in many countries and some international research programs to investigate the applicability to high burnup nuclear fuels have been performing as the existing safety criteria of DBA such as LOCA and RIA was established several decades ago. In this report, main research programs for the safety criteria of DBA such as LOCA and RIA are introduced, and also the current status on the modification of the safety criteria are also introduced.

  8. Advanced Corrosion-Resistant Zr Alloys for High Burnup and Generation IV Applications

    Energy Technology Data Exchange (ETDEWEB)

    Arthur Motta; Yong Hwan Jeong; R.J. Comstock; G.S. Was; Y.S. Kim

    2006-10-31

    The objective of this collaboration between four institutions in the US and Korea is to demonstrate a technical basis for the improvement of the corrosion resistance of zirconium-based alloys in more extreme operating environments (such as those present in severe fuel duty,cycles (high burnup, boiling, aggressive chemistry) andto investigate the feasibility (from the point of view of corrosion rate) of using advanced zirconium-based alloys in a supercritical water environment.

  9. Development of the CANDU high-burnup fuel design/analysis technology

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Ho Chun; Sim, K. S.; Oh, D. J.; Park, J. H.; Jun, J. S.; Yoo, K. J.

    1997-08-01

    This report contains all the information related to the development of the CANDU advanced fuel, so-called CANFLEX-NU, which is composed of 43 elements with natural uranium fuel. Also, it contains the compatibility study of CANFLEX-RU which is considered as a CANDU high burnup fuel. This report describes the mechanical design, thermalhydraulic and safety evaluations of CANFLEX fuel bundle. (author). 38 refs., 24 tabs., 74 figs.

  10. Analysis of the effect of UO{sub 2} high burnup microstructure on fission gas release

    Energy Technology Data Exchange (ETDEWEB)

    Jernkvist, Lars Olof; Massih, Ali [Quantum Technologies AB, Uppsala Science Park (Sweden)

    2002-10-01

    This report deals with high-burnup phenomena with relevance to fission gas release from UO{sub 2} nuclear fuel. In particular, we study how the fission gas release is affected by local buildup of fissile plutonium isotopes and fission products at the fuel pellet periphery, with subsequent formation of a characteristic high-burnup rim zone micro-structure. An important aspect of these high-burnup effects is the degradation of fuel thermal conductivity, for which prevalent models are analysed and compared with respect to their theoretical bases and supporting experimental data. Moreover, the Halden IFA-429/519.9 high-burnup experiment is analysed by use of the FRAPCON3 computer code, into which modified and extended models for fission gas release are introduced. These models account for the change in Xe/Kr-ratio of produced and released fission gas with respect to time and space. In addition, several alternative correlations for fuel thermal conductivity are implemented, and their impact on calculated fission gas release is studied. The calculated fission gas release fraction in IFA-429/519.9 strongly depends on what correlation is used for the fuel thermal conductivity, since thermal release dominates over athermal release in this particular experiment. The conducted calculations show that athermal release processes account for less than 10% of the total gas release. However, athermal release from the fuel pellet rim zone is presumably underestimated by our models. This conclusion is corroborated by comparisons between measured and calculated Xe/Kr-ratios of the released fission gas.

  11. S∧4 Reactor: Operating Lifetime and Estimates of Temperature and Burnup Reactivity Coefficients

    Science.gov (United States)

    King, Jeffrey C.; El-Genk, Mohamed S.

    2006-01-01

    The S∧4 reactor has a sectored, Mo-14%Re solid core for avoidance of single point failures in reactor cooling and Closed Brayton Cycle (CBC) energy conversion. The reactor is loaded with UN fuel, cooled with a He-Xe gas mixture at ~1200 K and operates at steady thermal power of 550 kW. Following a launch abort accident, the axial and radial BeO reflectors easily disassemble upon impact so that the bare reactor is subcriticial when submerged in wet sand or seawater and the core voids are filled with seawater. Spectral Shift Absorber (SSA) additives have been shown to increase the UN fuel enrichment and significantly reduce the total mass of the reactor. This paper investigates the effects of SSA additions on the temperature and burnup reactivity coefficients and the operational lifetime of the S∧4 reactor. SSAs slightly decrease the temperature reactivity feedback coefficient, but significantly increase the operating lifetime by decreasing the burnup reactivity coefficient. With no SSAs, fuel enrichment is only 58.5 wt% and the estimated operating lifetime is the shortest (7.6 years) with the highest temperature and burnup reactivity feedback coefficients (-0.2709 ¢/K and -1.3470 $/atom%). With europium-151 and gadolinium-155 additions, the enrichment (91.5 and 94 wt%) and operating lifetime (9.9 and 9.8 years) of the S∧4 reactor are the highest while the temperature and burnup reactivity coefficients (-0.2382 and -0.2447 ¢/K -0.9073 and 0.8502 $/atom%) are the lowest.

  12. Burn-up credit in criticality safety of PWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mahmoud, Rowayda F., E-mail: Rowayda_mahmoud@yahoo.com [Metallurgy Department, Nuclear Research Center, Atomic Energy Authority (Egypt); Shaat, Mohamed K. [Nuclear Engineering, Reactors Department, Nuclear Research Center, Atomic Energy Authority (Egypt); Nagy, M.E.; Agamy, S.A. [Professor of Nuclear Engineering, Nuclear and Radiation Department, Alexandria University (Egypt); Abdelrahman, Adel A. [Metallurgy Department, Nuclear Research Center, Atomic Energy Authority (Egypt)

    2014-12-15

    Highlights: • Designing spent fuel wet storage using WIMS-5D and MCNP-5 code. • Studying fresh and burned fuel with/out absorber like “B{sub 4}C and Ag–In–Cd” in racks. • Sub-criticality was confirmed for fresh and burned fuel under specific cases. • Studies for BU credit recommend increasing fuel burn-up to 60.0 GWD/MTU. • Those studies require new core structure materials, fuel composition and cladding. - Abstract: The criticality safety calculations were performed for a proposed design of a wet spent fuel storage pool. This pool will be used for the storage of spent fuel discharged from a typical pressurized water reactor (PWR). The mathematical model based on the international validated codes, WIMS-5 and MCNP-5 were used for calculating the effective multiplication factor, k{sub eff}, for the spent fuel stored in the pool. The data library for the multi-group neutron microscopic cross-sections was used for the cell calculations. The k{sub eff} was calculated for several changes in water density, water level, assembly pitch and burn-up with different initial fuel enrichment and new types and amounts of fixed absorbers. Also, k{sub eff} was calculated for the conservative fresh fuel case. The results of the calculations confirmed that the effective multiplication factor for the spent fuel storage is sub-critical for all normal and abnormal states. The future strategy for the burn-up credit recommends increasing the fuel burn-up to a value >60.0 GWD/MTU, which requires new fuel composition and new fuel cladding material with the assessment of the effects of negative reactivity build up.

  13. Evaluation of the Frequency for Gas Sampling for the High Burnup Confirmatory Data Project

    Energy Technology Data Exchange (ETDEWEB)

    Stockman, Christine T. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Alsaed, Halim A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Marschman, Steven C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Scaglione, John M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-05-01

    This report provides a technically based gas sampling frequency strategy for the High Burnup (HBU) Confirmatory Data Project. The evaluation of: 1) the types and magnitudes of gases that could be present in the project cask and, 2) the degradation mechanisms that could change gas compositions culminates in an adaptive gas sampling frequency strategy. This adaptive strategy is compared against the sampling frequency that has been developed based on operational considerations.

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

    Energy Technology Data Exchange (ETDEWEB)

    Saltzstein, Sylvia J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sorenson, Ken B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hanson, Brady [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Billone, Mike [Argonne National Lab. (ANL), Argonne, IL (United States); Scaglione, John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Montgomery, Rose [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bevard, Bruce [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-07-01

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

  15. Analysis of the growth of concomitant nitride layers produced by a post-discharge assisted process

    Energy Technology Data Exchange (ETDEWEB)

    Oseguera, J. [ITESM-CEM, Carretera al Lago de Guadalupe km. 3.5 Atizapan, 52926 (Mexico)]. E-mail: joseguer@itesm.mx; Castillo, F. [ITESM-CEM, Carretera al Lago de Guadalupe km. 3.5 Atizapan, 52926 (Mexico); Gomez, A. [UFRO, Av. Francisco Salazar 01145, Temuco, Casilla 54-d (Chile); Fraguela, A. [BUAP, Rio Verde y Ave. San Claudio, San Manuel, Puebla, 72570 (Mexico)

    2006-11-23

    In the present work, the growth of concomitant nitride layers during a post-discharge process is studied. The analysis takes into account the similarities and differences between nitriding post-discharge processes and other nitriding processes, employing a mathematical simulation of nitrogen diffusion. The considered differences are related to the thermodynamic standard states, the nitrogen concentration on the surface and the sputtering of the surface (this one for plasma processes). Nitrogen diffusion and layer formation are described from the beginning of the process by means of a mathematical model.

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

    Science.gov (United States)

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

    2014-06-01

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

  17. Needs of reliable nuclear data and covariance matrices for Burnup Credit in JEFF-3 library

    Directory of Open Access Journals (Sweden)

    Lecarpentier D.

    2013-03-01

    Full Text Available Burnup Credit (BUC is the concept which consists in taking into account credit for the reduction of nuclear spent fuel reactivity due to its burnup. In the case of PWR-MOx spent fuel, studies pointed out that the contribution of the 15 most absorbing, stable and non-volatile fission products selected to the credit is as important as the one of the actinides. In order to get a “best estimate” value of the keff, biases of their inventory calculation and individual reactivity worth should be considered in criticality safety studies. This paper enhances the most penalizing bias towards criticality and highlights possible improvements of nuclear data for the 15 FPs of PWRMOx BUC. Concerning the fuel inventory, trends in function of the burnup can be derived from experimental validation of the DARWIN-2.3 package (using the JEFF-3.1.1/SHEM library. Thanks to the BUC oscillation programme of separated FPs in the MINERVE reactor and fully validated scheme PIMS, calculation over experiment ratios can be accurately transposed to tendencies on the FPs integral cross sections.

  18. Thermal Hydraulic Analysis of 3 MW TRIGA Research Reactor of Bangladesh Considering Different Cycles of Burnup

    Directory of Open Access Journals (Sweden)

    M.H. Altaf

    2014-12-01

    Full Text Available Burnup dependent steady state thermal hydraulic analysis of TRIGA Mark-II research reactor has been carried out utilizing coupled point kinetics, neutronics and thermal hydraulics code EUREKA-2/RR. From the previous calculations of neutronics parameters including percentage burnup of individual fuel elements performed so far for 700 MWD burnt core of TRIGA reactor showed that the fuel rod predicted as hottest at the beginning of cycle (fresh core was found to remain as the hottest until 200 MWD of burn, but, with the progress of core burn, the hottest rod was found to be shifted and another rod in the core became the hottest. The present study intends to evaluate the thermal hydraulic parameters of these hottest fuel rods at different cycles of burnup, from beginning to 700 MWD core burnt considering reactor operates under steady state condition. Peak fuel centerline temperature, maximum cladding and coolant temperatures of the hottest channels were calculated. It revealed that maximum temperature reported for fuel clad and fuel centerline found to lie below their melting points which indicate that there is no chance of burnout on the fuel cladding surface and no blister in the fuel meat throughout the considered cycles of core burnt.

  19. Effect of Fuel Fraction on Small Modified CANDLE Burn-up Based Gas Cooled Fast Reactors

    Science.gov (United States)

    Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Khairurrijal, Asiah, Nur; Shafii, M. Ali

    2010-12-01

    A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE Burn-up has been performed. The objective of this research is to get optimal design parameters of such type reactors. The parameters of nuclear design including the critical condition, conversion ratio, and burn-up level were compared. These parameters are calculated by variation in the fuel fraction 47.5% up to 70%. Two dimensional full core multi groups diffusion calculations was performed by CITATION code. Group constant preparations are performed by using SRAC code system with JENDL-3.2 nuclear data library. In this design the reactor cores with cylindrical cell two dimensional R-Z core models are subdivided into several parts with the same volume in the axial directions. The placement of fuel in core arranged so that the result of plutonium from natural uranium can be utilized optimally for 10 years reactor operation. Modified CANDLE burn-up was established successfully in a core radial width 1.4 m. Total thermal power output for reference core is 550 MW. Study on the effect of fuel to coolant ratio shows that effective multiplication factor (keff) is in almost linear relations with the change of the fuel volume to coolant ratio.

  20. Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation - Vandellos II Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Gauld, Ian C [ORNL

    2011-01-01

    This report is one of the several recent NUREG/CR reports documenting benchmark-quality radiochemical assay data and the use of the data to validate computer code predictions of isotopic composition for spent nuclear fuel, to establish the uncertainty and bias associated with code predictions. The experimental data analyzed in the current report were acquired from a high-burnup fuel program coordinated by Spanish organizations. The measurements included extensive actinide and fission product data of importance to spent fuel safety applications, including burnup credit, decay heat, and radiation source terms. Six unique spent fuel samples from three uranium oxide fuel rods were analyzed. The fuel rods had a 4.5 wt % {sup 235}U initial enrichment and were irradiated in the Vandellos II pressurized water reactor operated in Spain. The burnups of the fuel samples range from 42 to 78 GWd/MTU. The measurements were used to validate the two-dimensional depletion sequence TRITON in the SCALE computer code system.

  1. Oxygen potential in the rim region of high burnup UO 2 fuel

    Science.gov (United States)

    Matzke, Hj.

    1994-01-01

    Small specimens from the rim region (fuel surface) of a UO 2 fuel rod with an average burnup of 7.6% FIMA were analysed in a miniaturized galvanic cell to determine their oxygen potential Δ Ḡ(O 2) . These fuel pieces represented the porous rim structure with very small grains known to be formed near the periphery of high burnup UO 2 fuel pellets. The oxygen potential of the rim material was very low, corresponding to that of unirradiated stoichiometric UO 2, or to that of slightly substoichiometric UO 2 containing rare earth fission products. No indication of oxidation due to fission was found, though most fission was that of Pu. Measurements on pieces from the inner, unrestructured fuel showed somewhat higher oxygen potentials corresponding to those of very slightly substoichiometric fuel if allowance is made for the incorporation of rare earths. These results are in contrast to some generally accepted ideas of burnup effects, and the possible reasons and implications are discussed.

  2. Heterostructures for Increased Quantum Efficiency in Nitride LEDs

    Energy Technology Data Exchange (ETDEWEB)

    Davis, Robert F. [Carnegie Mellon Univ., Pittsburgh, PA (United States)

    2010-09-30

    Task 1. Development of an advanced LED simulator useful for the design of efficient nitride-based devices. Simulator will contain graphical interface software that can be used to specify the device structure, the material parameters, the operating conditions and the desired output results. Task 2. Theoretical and experimental investigations regarding the influence on the microstructure, defect concentration, mechanical stress and strain and IQE of controlled changes in the chemistry and process route of deposition of the buffer layer underlying the active region of nitride-based blue- and greenemitting LEDs. Task 3. Theoretical and experimental investigations regarding the influence on the physical properties including polarization and IQE of controlled changes in the geometry, chemistry, defect density, and microstructure of components in the active region of nitride-based blue- and green-emitting LEDs. Task 4. Theoretical and experimental investigations regarding the influence on IQE of novel heterostructure designs to funnel carriers into the active region for enhanced recombination efficiency and elimination of diffusion beyond this region. Task 5. Theoretical and experimental investigations regarding the influence of enhanced p-type doping on the chemical, electrical, and microstructural characteristics of the acceptor-doped layers, the hole injection levels at Ohmic contacts, the specific contact resistivity and the IQE of nitride-based blue- and green-emitting LEDs. Development and optical and electrical characterization of reflective Ohmic contacts to n- and p-type GaN films.

  3. Analytical and Experimental Evaluation of Joining Silicon Carbide to Silicon Carbide and Silicon Nitride to Silicon Nitride for Advanced Heat Engine Applications Phase II

    Energy Technology Data Exchange (ETDEWEB)

    Sundberg, G.J.

    1994-01-01

    Techniques were developed to produce reliable silicon nitride to silicon nitride (NCX-5101) curved joins which were used to manufacture spin test specimens as a proof of concept to simulate parts such as a simple rotor. Specimens were machined from the curved joins to measure the following properties of the join interlayer: tensile strength, shear strength, 22 C flexure strength and 1370 C flexure strength. In parallel, extensive silicon nitride tensile creep evaluation of planar butt joins provided a sufficient data base to develop models with accurate predictive capability for different geometries. Analytical models applied satisfactorily to the silicon nitride joins were Norton's Law for creep strain, a modified Norton's Law internal variable model and the Monkman-Grant relationship for failure modeling. The Theta Projection method was less successful. Attempts were also made to develop planar butt joins of siliconized silicon carbide (NT230).

  4. 纯钛材离子渗氮后在Hank's人工模拟体液中的电化学性能%Electrochemical Behavior and Corrosion Resistance of Ion Nitriding Layer on Titanium Substrate in Artificially Simulated Body Solution

    Institute of Scientific and Technical Information of China (English)

    吴新春; 陈吉; 韩啸

    2013-01-01

    As-cast Ti substrate was ion nitridized so as to increase the surface hardness and corrosion resistance as well as compatibility to human body. The elemental composition, phase structure, and morphology of as - obtained ion nitriding layer were analyzed with an energy dispersive spectrometer, an X-ray diffractometer, and a scanning electron microscope, while its hardness was measured with a hardness meter. Moreover, the electrochemical behavior and corrosion resistance of as - cast Ti substrate and the ion nitriding layer in Hank' s solution of 37 ℃ were comparatively investigated based on measurements of potentiodynamic polarization curves and alternating current impedance spectra. It was found that a gradient ion nitriding layer with a thickness of 17 μm was formed on Ti substrate surface after ion nitriding at 650 ℃. As -obtained ion nitriding layer consisted of α - Ti ( N) , Ti2 N and TiN. Moreover, the ion nitriding layer had a hardness of about (650 ±20) HV, which was 2. 3 times as much as that of the Ti substrate, while the Ti substrate after ion nitriding possessed greatly improved corrosion resistance.%为了提高纯钛材的表面硬度及耐蚀性,使之更适应人体环境,对铸态纯钛材表面进行离子渗氮.利用能谱仪、X射线衍射仪、扫描电镜及显微硬度计研究了渗氮钛材的成分、相结构、形貌及硬度,利用动电位极化曲线及交流阻抗谱对比研究了纯钛材及离子渗氮钛材在37℃的Hank's人工模拟体液中的电化学行为.结果表明:经过650℃渗氮处理后,铸态纯钛材表面形成了17 μm的梯度渗氮层,其相组成为α-Ti(N),Ti2N,TiN;表面显微硬度约为(650±20) HV,为铸态纯钛材的2.3倍;离子渗氮钛材的耐蚀性明显提高.

  5. Structural properties of iron nitride on Cu(100): An ab-initio molecular dynamics study

    KAUST Repository

    Heryadi, Dodi

    2011-01-01

    Due to their potential applications in magnetic storage devices, iron nitrides have been a subject of numerous experimental and theoretical investigations. Thin films of iron nitride have been successfully grown on different substrates. To study the structural properties of a single monolayer film of FeN we have performed an ab-initio molecular dynamics simulation of its formation on a Cu(100) substrate. The iron nitride layer formed in our simulation shows a p4gm(2x2) reconstructed surface, in agreement with experimental results. In addition to its structural properties, we are also able to determine the magnetization of this thin film. Our results show that one monolayer of iron nitride on Cu(100) is ferromagnetic with a magnetic moment of 1.67 μ B. © 2011 Materials Research Society.

  6. Leachability of nitrided ilmenite in hydrochloric acid

    CSIR Research Space (South Africa)

    Swanepoel, JJ

    2010-10-01

    Full Text Available Titanium nitride in upgraded nitrided ilmenite (bulk of iron removed) can selectively be chlorinated to produce titanium tetrachloride. Except for iron, most other components present during this low temperature (ca. 200 °C) chlorination reaction...

  7. Plasmonic titanium nitride nanostructures for perfect absorbers

    DEFF Research Database (Denmark)

    Guler, Urcan; Li, Wen-Wei; Kinsey, Nathaniel

    2013-01-01

    We propose a metamaterial based perfect absorber in the visible region, and investigate the performance of titanium nitride as an alternative plasmonic material. Numerical and experimental results reveal that titanium nitride performs better than gold as a plasmonic absorbing material...

  8. Cathodic Cage Plasma Nitriding: An Innovative Technique

    OpenAIRE

    Sousa,R.R.M.; de Araújo, F. O.; J. A. P. da Costa; Brandim,A.S.; R. A. de Brito; C. Alves

    2012-01-01

    Cylindrical samples of AISI 1020, AISI 316, and AISI 420 steels, with different heights, were simultaneously treated by a new technique of ionic nitriding, entitled cathodic cage plasma nitriding (CCPN), in order to evaluate the efficiency of this technique to produce nitrided layers with better properties compared with those obtained using conventional ionic nitriding technique. This method is able to eliminate the edge effect in the samples, promoting a better uniformity of temperature, and...

  9. A comparative study of MONTEBURNS and MCNPX 2.6.0 codes in ADS simulations

    Energy Technology Data Exchange (ETDEWEB)

    Barros, Graiciany P.; Pereira, Claubia; Veloso, Maria A.F.; Velasquez, Carlos E.; Costa, Antonella L., E-mail: gbarros@ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2013-07-01

    The possible use of the MONTEBURNS and MCNPX 2.6.0 codes in Accelerator-driven systems (ADSs) simulations for fuel evolution description is discussed. ADSs are investigated for fuel breeding and long-lived fission product transmutation so simulations of fuel evolution have a great relevance. The burnup/depletion capability is present in both studied codes. MONTEBURNS code links Monte Carlo N-Particle Transport Code (MCNP) to the radioactive decay burnup code ORIGEN2, whereas MCNPX depletion/ burnup capability is a linked process involving steady-state flux calculations by MCNPX and nuclide depletion calculations by CINDER90. A lead-cooled accelerator-driven system fueled with thorium was simulated and the results obtained using MONTEBURNS code and the results from MCNPX 2.6.0 code were compared. The system criticality and the variation of the actinide inventory during the burnup were evaluated and the results indicate a similar behavior between the results of each code. (author)

  10. Theoretical Compton profile of diamond, boron nitride and carbon nitride

    Science.gov (United States)

    Aguiar, Julio C.; Quevedo, Carlos R.; Gomez, José M.; Di Rocco, Héctor O.

    2017-09-01

    In the present study, we used the generalized gradient approximation method to determine the electron wave functions and theoretical Compton profiles of the following super-hard materials: diamond, boron nitride (h-BN), and carbon nitride in its two known phases: βC3N4 and gC3N4 . In the case of diamond and h-BN, we compared our theoretical results with available experimental data. In addition, we used the Compton profile results to determine cohesive energies and found acceptable agreement with previous experiments.

  11. Review of Technical Issues Related to Predicting Isotopic Compositions and Source Terms for High-Burnup LWR Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I. C.; Parks, C. V.

    2000-12-11

    This report has been prepared to review the technical issues important to the prediction of isotopic compositions and source terms for high-burnup, light-water-reactor (LWR) fuel as utilized in the licensing of spent fuel transport and storage systems. The current trend towards higher initial 235U enrichments, more complex assembly designs, and more efficient fuel management schemes has resulted in higher spent fuel burnups than seen in the past. This trend has led to a situation where high-burnup assemblies from operating LWRs now extend beyond the area where available experimental data can be used to validate the computational methods employed to calculate spent fuel inventories and source terms. This report provides a brief review of currently available validation data, including isotopic assays, decay heat measurements, and shielded dose-rate measurements. Potential new sources of experimental data available in the near term are identified. A review of the background issues important to isotopic predictions and some of the perceived technical challenges that high-burnup fuel presents to the current computational methods are discussed. Based on the review, the phenomena that need to be investigated further and the technical issues that require resolution are presented. The methods and data development that may be required to address the possible shortcomings of physics and depletion methods in the high-burnup and high-enrichment regime are also discussed. Finally, a sensitivity analysis methodology is presented. This methodology is currently being investigated at the Oak Ridge National Laboratory as a computational tool to better understand the changing relative significance of the underlying nuclear data in the different enrichment and burnup regimes and to identify the processes that are dominant in the high-burnup regime. The potential application of the sensitivity analysis methodology to help establish a range of applicability for experimental

  12. Mathematical Modelling of Nitride Layer Growth of Low Temperature Gas and Plasma Nitriding of AISI 316L

    Directory of Open Access Journals (Sweden)

    Triwiyanto A.

    2014-07-01

    Full Text Available This paper present mathematical model which developed to predict the nitrided layer thickness (case depth of gas nitrided and plasma nitrided austenitic stainless steel according to Fick’s first law for pure iron by adapting and manipulating the Hosseini’s model to fit the diffusion mechanism where nitrided structure formed by nitrided AISI 316L austenitic stainless steel. The mathematical model later tested against various actual gas nitriding and plasma nitriding experimental results with varying nitriding temperature and nitriding duration to see whether the model managed to successfully predict the nitrided layer thickness. This model predicted the coexistence of ε-Fe2-3N and γ΄-Fe4N under the present nitriding process parameters. After the validation process, it is proven that the mathematical model managed to predict the nitrided layer growth of the gas nitrided and plasma nitrided of AISI 316L SS up to high degree of accuracy.

  13. A criticality analysis of the GBC-32 dry storage cask with Hanbit nuclear power plant unit 3 fuel assemblies from the viewpoint of burnup credit

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Hyung Ju; Kim, Do Yeon; Park, Kwang Heon; Hong, Ser Gi [Dept. of Nuclear Engineering, Kyung Hee University, Seoul (Korea, Republic of)

    2016-06-15

    Nuclear criticality safety analyses (NCSAs) considering burnup credit were performed for the GBC-32 cask. The used nuclear fuel assemblies (UNFAs) discharged from Hanbit Nuclear Power Plant Unit 3 Cycle 6 were loaded into the cask. Their axial burnup distributions and average discharge burnups were evaluated using the DeCART and Multi-purpose Analyzer for Static and Transient Effects of Reactors (MASTER) codes, and NCSAs were performed using SCALE 6.1/STandardized Analysis of Reactivity for Burnup Credit using SCALE (STARBUCS) and Monte Carlo N-Particle transport code, version 6 (MCNP 6). The axial burnup distributions were determined for 20 UNFAs with various initial enrichments and burnups, which were applied to the criticality analysis for the cask system. The UNFAs for 20- and 30-year cooling times were assumed to be stored in the cask. The criticality analyses indicated that keff values for UNFAs with nonuniform axial burnup distributions were larger than those with a uniform distribution, that is, the end effects were positive but much smaller than those with the reference distribution. The axial burnup distributions for 20 UNFAs had shapes that were more symmetrical with a less steep gradient in the upper region than the reference ones of the United States Department of Energy. These differences in the axial burnup distributions resulted in a significant reduction in end effects compared with the reference.

  14. III-Nitride nanowire optoelectronics

    Science.gov (United States)

    Zhao, Songrui; Nguyen, Hieu P. T.; Kibria, Md. G.; Mi, Zetian

    2015-11-01

    Group-III nitride nanowire structures, including GaN, InN, AlN and their alloys, have been intensively studied in the past decade. Unique to this material system is that its energy bandgap can be tuned from the deep ultraviolet (~6.2 eV for AlN) to the near infrared (~0.65 eV for InN). In this article, we provide an overview on the recent progress made in III-nitride nanowire optoelectronic devices, including light emitting diodes, lasers, photodetectors, single photon sources, intraband devices, solar cells, and artificial photosynthesis. The present challenges and future prospects of III-nitride nanowire optoelectronic devices are also discussed.

  15. Boron Nitride Nanoribbons from Exfoliation of Boron Nitride Nanotubes

    Science.gov (United States)

    Hung, Ching-Cheh; Hurst, Janet; Santiago, Diana

    2017-01-01

    Two types of boron nitride nanotubes (BNNTs) were exfoliated into boron nitride nanoribbons (BNNR), which were identified using transmission electron microscopy: (1) commercial BNNTs with thin tube walls and small diameters. Tube unzipping was indicated by a large decrease of the sample's surface area and volume for pores less than 2 nm in diameter. (2) BNNTs with large diameters and thick walls synthesized at NASA Glenn Research Center. Here, tube unraveling was indicated by a large increase in external surface area and pore volume. For both, the exfoliation process was similar to the previous reported method to exfoliate commercial hexagonal boron nitride (hBN): Mixtures of BNNT, FeCl3, and NaF (or KF) were sequentially treated in 250 to 350 C nitrogen for intercalation, 500 to 750 C air for exfoliation, and finally HCl for purification. Property changes of the nanosized boron nitride throughout this process were also similar to the previously observed changes of commercial hBN during the exfoliation process: Both crystal structure (x-ray diffraction data) and chemical properties (Fourier-transform infrared spectroscopy data) of the original reactant changed after intercalation and exfoliation, but most (not all) of these changes revert back to those of the reactant once the final, purified products are obtained.

  16. Homogeneous dispersion of gallium nitride nanoparticles in a boron nitride matrix by nitridation with urea.

    Science.gov (United States)

    Kusunose, Takafumi; Sekino, Tohru; Ando, Yoichi

    2010-07-01

    A Gallium Nitride (GaN) dispersed boron nitride (BN) nanocomposite powder was synthesized by heating a mixture of gallium nitrate, boric acid, and urea in a hydrogen atmosphere. Before heat treatment, crystalline phases of urea, boric acid, and gallium nitrate were recognized, but an amorphous material was produced by heat treatment at 400 degrees C, and then was transformed into GaN and turbostratic BN (t-BN) by further heat treatment at 800 degrees C. TEM obsevations of this composite powder revealed that single nanosized GaN particles were homogeneously dispersed in a BN matrix. Homogeneous dispersion of GaN nanoparticles was thought to be attained by simultaneously nitriding gallium nitrate and boric acid to GaN and BN with urea.

  17. Theoretical treatment of nitriding and nitrocarburizing of iron

    Science.gov (United States)

    Du, Hong; Ågren, John

    1996-04-01

    Mathematical models are developed for both nitriding and nitrocarburizing of iron taking into account the diffusion of N or C and N through various phases and the thermodynamic properties of the ternary Fe-C-N system. Analytical solutions are obtained for the ɛ/γ' bilayer growth of the compound layer assuming constant diffusion coefficients, and the results are compared with those obtained from numerical simulations taking into account the concentration-dependent diffusivities. No significant difference was found between these two methods for nitriding of iron. For nitrocarburizing of iron, it was found that the off-diagonal diffusivities of the ɛ and γ' phases must be taken into account in the analytical solution in order to obtain reasonable results. In addition, it is shown that the phase constitution of the compound layer produced during nitrocarburizing of iron can be predicted by the numerical simulation.

  18. Use of burnup credit in criticality evaluation for spent fuel storage pool

    Energy Technology Data Exchange (ETDEWEB)

    Chon, Je Keun; Kim, Jae Chun; Koh, Duck Joon; Kim Byung Tae [Nuclear Environment Technology Institute, Korea Electric Power Corporation, Taejon (Korea, Republic of)

    1999-07-01

    Boraflex is a polymer based material which is used as matrix to contain a neutron absorber material, boron carbide. In a typical spent fuel pool the irradiated Boraflex has been known as a significant source of silica. Since 1996, it was reported that elevated silica levels were measured in the Ulchin Unit 2 spent fuel pool water. Therefore, the Ulchin Unit 2 spent fuel storage racks were needed to be reanalyzed to allow storage of fuel assemblies with normal enrichments up to 5.0w/o U-235 in all storage cell locations using credit for burnup. The analysis does not take any credit for the presence of the spent fuel rack Boraflex neutron absorber panels. In region 2, the calculations were performed by assuming in an infinite radial array of storage cells. No credit is taken for axial or radial neutron leakage. The water in the spent fuel storage pool was assumed to be pure. In the evaluation of the Ulchin Unit 2 spent fuel storage pool, criticality analyses were performed with the CASMO-3 code. A reactivity uncertainty in the fuel depletion calculations was combined with other calculational uncertainty. The manufacturing tolerances were considered, as well. From the calculation, the acceptable burnup domain in region 2 of the spent fuel storage pool. where the curve identifies conditions of equal reactivity for various initial enrichments between 1.6w/o and 5.0w/o, was evaluated. In region 2, the maximum k{sub e}ff including all uncertainties, is 0.94648 for the enrichment-burnup combination from loading curve. (author)

  19. Application of SCALE4.4 system for burnup credit criticality analysis of PWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Hee Sung; Ro, Seung gy; Bae, Kang mok; Shin, YoungJoon; Kim, Ik Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-07-01

    An investigation on the application of burnup credit for a PWR spent fuel storage pool has been carried out with the use of the SCALE 4.4 computer code system consisting of SAS2H and CSAS6 modules in association with 44-group SCALE cross-section library. Prior to the application of the computer code system, a series of bench markings have been performed in comparison with available data. A benchmarking of the SAS2h module has been done for experimental concentration data of 54 PWR spent fuel and then correction factors with a 95% probability at a 95% confidence level have been determined on the basis of the calculated and measured concentrations of 38 nuclides. After that, the bias which might have resulted from the use of the CSAS6 module has been calculated for 46 criticality experimental data of UO{sub 2} fuel and MOX fuel assemblies. The calculation bias with one-sided tolerance limit factor (2.086) corresponding to a 95% probability at a 95% confidence level has consequently been obtained to be 0.00834. Burnup credit criticality analysis has been done for the PWR spent fuel storage pool by means of the benchmarked or validated code system. It is revealed that the minimum burnup for safe storage is 7560 MWd/tU in 5 wt% enriched fuel if both actinides and fission products in spent fuel are taken into account. However, the minimum value required seems to be 9,565 MWd/tU in the same enriched fuel provided that only the actinides are taken into consideration. (author)

  20. Spent fuel dissolution rates as a function of burnup and water chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Gray, W.J.

    1998-06-01

    To help provide a source term for performance-assessment calculations, dissolution studies on light-water-reactor (LWR) spent fuel have been conducted over the past few years at Pacific Northwest National Laboratory in support of the Yucca Mountain Site Characterization Project. This report describes that work for fiscal years 1996 through mid-1998 and includes summaries of some results from previous years for completeness. The following conclusions were based on the results of various flowthrough dissolution rate tests and on tests designed to measure the inventories of {sup 129}I located within the fuel/cladding gap region of different spent fuels: (1) Spent fuels with burnups in the range 30 to 50 MWd/kgM all dissolved at about the same rate over the conditions tested. To help determine whether the lack of burnup dependence extends to higher and lower values, tests are in progress or planned for spent fuels with burnups of 13 and {approximately} 65 MWd/kgM. (2) Oxidation of spent fuel up to the U{sub 4}O{sub 9+x} stage does not have a large effect on intrinsic dissolution rates. However, this degree of oxidation could increase the dissolution rates of relatively intact fuel by opening the grain boundaries, thereby increasing the effective surface area that is available for contact by water. From a disposal viewpoint, this is a potentially more important consideration than the effect on intrinsic rates. (3) The gap inventories of {sup 129}I were found to be smaller than the fission gas release (FGR) for the same fuel rod with the exception of the rod with the highest FGR. Several additional fuels would have to be tested to determine whether a generalized relationship exists between FGR and {sup 129}I gap inventory for US LWR fuels.

  1. Burnup concept for a long-life fast reactor core using MCNPX.

    Energy Technology Data Exchange (ETDEWEB)

    Holschuh, Thomas Vernon,; Lewis, Tom Goslee,; Parma, Edward J.,

    2013-02-01

    This report describes a reactor design with a burnup concept for a long-life fast reactor core that was evaluated using Monte Carlo N-Particle eXtended (MCNPX). The current trend in advanced reactor design is the concept of a small modular reactor (SMR). However, very few of the SMR designs attempt to substantially increase the lifetime of a reactor core, especially without zone loading, fuel reshuffling, or other artificial mechanisms in the core that %E2%80%9Cflatten%E2%80%9D the power profile, including non-uniform cooling, non-uniform moderation, or strategic poison placement. Historically, the limitations of computing capabilities have prevented acceptable margins in the temporal component of the spatial excess reactivity in a reactor design, due primarily to the error in burnup calculations. This research was performed as an initial scoping analysis into the concept of a long-life fast reactor. It can be shown that a long-life fast reactor concept can be modeled using MCNPX to predict burnup and neutronics behavior. The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional Light Water Reactors (LWRs) or other SMR designs. For the purpose of this study, a single core design was investigated: a relatively small reactor core, yielding a medium amount of power (~200 to 400 MWth). The results of this scoping analysis were successful in providing a preliminary reactor design involving metal U-235/U-238 fuel with HT-9 fuel cladding and sodium coolant at a 20% volume fraction.

  2. Criticality Analysis of Assembly Misload in a PWR Burnup Credit Cask

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J. C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2008-01-31

    The Interim Staff Guidance on bumup credit (ISG-8) for spent fuel in storage and transportation casks, issued by the Nuclear Regulatory Commission's Spent Fuel Project Office, recommends a bumup measurement for each assembly to confirm the reactor record and compliance with the assembly bumup value used for loading acceptance. This recommendation is intended to prevent unauthorized loading (misloading) of assemblies due to inaccuracies in reactor burnup records and/or improper assembly identification, thereby ensuring that the appropriate subcritical margin is maintained. This report presents a computational criticality safety analysis of the consequences of misloading fuel assemblies in a highcapacity cask that relies on burnup credit for criticality safety. The purpose of this report is to provide a quantitative understanding of the effects of fuel misloading events on safety margins. A wide variety of fuel-misloading configurations are investigated and results are provided for informational purposes. This report does not address the likelihood of occurrence for any of the misload configurations considered. For representative, qualified bumup-enrichment combinations, with and without fission products included, misloading two assemblies that are underburned by 75% results in an increase in keff of 0.025-0.045, while misloading four assemblies that are underburned by 50% also results in an increase in keff of 0.025-0.045. For the cask and conditions considered, a reduction in bumup of 20% in all assemblies results in an increase in kff of less than 0.035. Misloading a single fresh assembly with 3, 4, or 5 wt% 235U enrichment results in an increase in keffof--0.02, 0.04, or 0.06, respectively. The report concludes with a summary of these and other important findings, as well as a discussion of relevant issues that should be considered when assessing the appropriate role of burnup measurements.

  3. Nanoporous Carbon Nitride: A High Efficient Filter for Seawater Desalination

    OpenAIRE

    Weifeng LI; Yang, Yanmei; Zhou, Hongcai; Zhang, Xiaoming; Zhao, Mingwen

    2015-01-01

    The low efficiency of commercially-used reverse osmosis (RO) membranes has been the main obstacle in seawater desalination application. Here, we report the auspicious performance, through molecular dynamics simulations, of a seawater desalination filter based on the recently-synthesized graphene-like carbon nitride (g-C2N) [Nat. Commun., 2015, 6, 6486]. Taking advantage of the inherent nanopores and excellent mechanical properties of g-C2N filter, highly efficient seawater desalination can be...

  4. Observation of viscoelasticity in boron nitride nanosheet aerogel.

    Science.gov (United States)

    Zeng, Xiaoliang; Ye, Lei; Sun, Rong; Xu, Jianbin; Wong, Ching-Ping

    2015-07-14

    The viscoelasticity of boron nitride nanosheet (BNNS) aerogel has been observed and investigated. It is found that the BNNS aerogel has a high damping ratio (0.2), while it exhibits lightweight and negligible temperature dependence below 180 °C. The creep behavior of the BNNS aerogel markedly demonstrates its strain dependence on stress magnitude and temperature, and can be well simulated by the classical models.

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

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Burtseva, T. A. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-08-30

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

  6. Thermodynamic analysis for high burn-up fuel internal chemistry. 2

    Energy Technology Data Exchange (ETDEWEB)

    Fuji, Kensho; Kyoh, Bunkei [Kinki Univ., Higashi-Osaka, Osaka (Japan)

    1998-09-01

    Thermodynamic calculations with the computer program SOLGASMIX-PV have been performed for the chemical states expected in irradiated fast breeder reactor (FBR) fuels containing transuranium (TRU) elements. The analysis shows that A (alkali and alkaline-earth)-molybdates exist, but neither A-uranates nor A-zirconates are formed in FBR fuel pellets irradiated to high burn-up. And increase of oxygen potential in irradiated FBR fuel is ascribed to growing amount of rare earth, noble metal and TRU elements. (author)

  7. Investigation into nitrided spur gears

    Energy Technology Data Exchange (ETDEWEB)

    Yilbas, B.S.; Coban, A.; Nickel, J.; Sunar, M.; Sami, M.; Abdul Aleem, B.J. [King Fahd Univ. of Petroleum and Minerals, Dhahran (Saudi Arabia)

    1996-12-01

    The cold forging method has been widely used in industry to produce machine parts. In general, gears are produced by shaping or hobbing. One of the shaping techniques is precision forging, which has several advantages over hobbing. In the present study, cold forging of spur gears from Ti-6Al-4V material is introduced. To improve the surface properties of the resulting gears, plasma nitriding was carried out. Nuclear reaction analysis was carried out to obtain the nitrogen concentration, while the micro-PIXE technique was used to determine the elemental distribution in the matrix after forging and nitriding processes. Scanning electron microscopy and x-ray powder diffraction were used to investigate the metallurgical changes and formation of nitride components in the surface region. Microhardness and friction tests were carried out to measure the hardness depth profile and friction coefficient at the surface. Finally, scoring failure tests were conducted to determine the rotational speed at which the gears failed. Three distinct regions were obtained in the nitride region, and at the initial stages of the scoring tests, failure in surface roughness was observed in the vicinity of the tip of the gear tooth. This occurred at a particular rotational speed and work input.

  8. Investigation into nitrided spur gears

    Science.gov (United States)

    Yilbas, B. S.; Coban, A.; Nickel, J.; Sunar, M.; Sami, M.; Aleem, B. J. Abdul

    1996-12-01

    The cold forging method has been widely used in industry to produce machine parts. In general, gears are produced by shaping or hobbing. One of the shaping techniques is precision forging, which has several advantages over hobbing. In the present study, cold forging of spur gears from Ti-6A1-4V material is introduced. To improve the surface properties of the resulting gears, plasma nitriding was carried out. Nuclear reaction analysis was carried out to obtain the nitrogen concentration, while the micro-PIXE technique was used to determine the elemental distribution in the matrix after forging and nitriding processes. Scanning electron microscopy and x-ray powder diffraction were used to investigate the metallurgical changes and formation of nitride components in the surface region. Microhardness and friction tests were carried out to measure the hardness depth profile and friction coefficient at the surface. Finally, scoring failure tests were conducted to determine the rotational speed at which the gears failed. Three distinct regions were obtained in the nitride region, and at the initial stages of the scoring tests, failure in surface roughness was observed in the vicinity of the tip of the gear tooth. This occurred at a particular rotational speed and work input.

  9. A semi-empirical model for the formation and depletion of the high burnup structure in UO2

    Science.gov (United States)

    Pizzocri, D.; Cappia, F.; Luzzi, L.; Pastore, G.; Rondinella, V. V.; Van Uffelen, P.

    2017-04-01

    In the rim zone of UO2 nuclear fuel pellets, the combination of high burnup and low temperature drives a microstructural change, leading to the formation of the high burnup structure (HBS). In this work, we propose a semi-empirical model to describe the formation of the HBS, which embraces the polygonisation/recrystallization process and the depletion of intra-granular fission gas, describing them as inherently related. For this purpose, we performed grain-size measurements on samples at radial positions in which the restructuring was incomplete. Based on these new experimental data, we infer an exponential reduction of the average grain size with local effective burnup, paired with a simultaneous depletion of intra-granular fission gas driven by diffusion. The comparison with currently used models indicates the applicability of the herein developed model within integral fuel performance codes.

  10. Thermal tuners on a Silicon Nitride platform

    CERN Document Server

    Pérez, Daniel; Baños, Rocío; Doménech, José David; Sánchez, Ana M; Cirera, Josep M; Mas, Roser; Sánchez, Javier; Durán, Sara; Pardo, Emilio; Domínguez, Carlos; Pastor, Daniel; Capmany, José; Muñoz, Pascual

    2016-01-01

    In this paper, the design trade-offs for the implementation of small footprint thermal tuners on silicon nitride are presented, and explored through measurements and supporting simulations of a photonic chip based on Mach-Zehnder Interferometers. Firstly, the electrical properties of the tuners are assessed, showing a compromise between compactness and deterioration. Secondly, the different variables involved in the thermal efficiency, switching power and heater dimensions, are analysed. Finally, with focus on exploring the limits of this compact tuners with regards to on chip component density, the thermal-cross talk is also investigated. Tuners with footprint of 270x5 {\\mu}m 2 and switching power of 350 mW are reported, with thermal-cross talk, in terms of induced phase change in adjacent devices of less than one order of magnitude at distances over 20 {\\mu}m. Paths for the improvement of thermal efficiency, power consumption and resilience of the devices are also outlined

  11. Research and verification of Monte Carlo burnup calculations based on Chebyshev rational approximation method%基于切比雪夫有理逼近方法的蒙特卡罗燃耗计算研究与验证

    Institute of Scientific and Technical Information of China (English)

    范文玎; 孙光耀; 张彬航; 陈锐; 郝丽娟

    2016-01-01

    燃耗计算在反应堆设计、分析研究中起着重要作用.相比于传统点燃耗算法,切比雪夫有理逼近方法(Chebyshev rational approximation method,CRAM)具有计算速度快、精度高的优点.基于超级蒙特卡罗核计算仿真软件系统SuperMC(Super Monte Carlo Simulation Program for Nuclear and Radiation Process),采用切比雪夫有理逼近方法和桶排序能量查找方法,进行了蒙特卡罗燃耗计算的初步研究与验证.通过燃料棒燃耗例题以及IAEA-ADS(International Atomic Energy Agency-Accelerator Driven Systems)国际基准题,初步验证了该燃耗计算方法的正确性,且IAEA-ADS基准题测试表明,与统一能量网格方法相比,桶排序能量查找方法在保证了计算效率的同时减少了内存开销.%Background:Burnup calculation is the key point of reactor design and analysis. It's significant to calculate the burnup situation and isotopic atom density accurately while a reactor is being designed.Purpose:Based on the Monte Carlo particle simulation code SuperMC (Super Monte Carlo Simulation Program for Nuclear and Radiation Process), this paper aimed to conduct preliminary study and verification on Monte Carlo burnup calculations. Methods:For the characteristics of accuracy, this paper adopted Chebyshev rational approximation method (CRAM) as the point-burnup algorithm. Moreover, instead of the union energy grids method, this paper adopted an energy searching method based on bucket sort algorithm, which reduced the memory overhead on the condition that the calculation efficiency is ensured.Results:By calculating the fuel rod burnup problem and the IAEA-ADS (International Atomic Energy Agency - Accelerator Driven Systems) international benchmark, the simulation results were basically consistent with Serpent and other counties' results, respectively. In addition, the bucket sort energy searching method reduced about 95% storage space compared with union energy grids method for IAEA

  12. EBSD and TEM characterization of high burn-up mixed oxide fuel

    Science.gov (United States)

    Teague, Melissa; Gorman, Brian; Miller, Brandon; King, Jeffrey

    2014-01-01

    Understanding and studying the irradiation behavior of high burn-up oxide fuel is critical to licensing of future fast breeder reactors. Advancements in experimental techniques and equipment are allowing for new insights into previously irradiated samples. In this work dual column focused ion beam (FIB)/scanning electron microscope (SEM) was utilized to prepared transmission electron microscope samples from mixed oxide fuel with a burn-up of 6.7% FIMA. Utilizing the FIB/SEM for preparation resulted in samples with a dose rate of <0.5 mRem/h compared to ∼1.1 R/h for a traditionally prepared TEM sample. The TEM analysis showed that the sample taken from the cooler rim region of the fuel pellet had ∼2.5× higher dislocation density than that of the sample taken from the mid-radius due to the lower irradiation temperature of the rim. The dual column FIB/SEM was additionally used to prepared and serially slice ∼25 μm cubes. High quality electron back scatter diffraction (EBSD) were collected from the face at each step, showing, for the first time, the ability to obtain EBSD data from high activity irradiated fuel.

  13. Evaluation of the HTTR criticality and burnup calculations with continuous-energy and multigroup cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Chiang, Min-Han; Wang, Jui-Yu [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Sheu, Rong-Jiun, E-mail: rjsheu@mx.nthu.edu.tw [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Liu, Yen-Wan Hsueh [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China)

    2014-05-01

    The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects.

  14. Void effect analysis of Pb-208 of fast reactors with modified CANDLE burn-up scheme

    Science.gov (United States)

    Widiawati, Nina; Su'ud, Zaki

    2015-09-01

    Void effect analysis of Pb-208 as coolant of fast reactors with modified candle burn-up scheme has been conducted. Lead cooled fast reactor (LFR) is one of the fourth-generation reactor designs. The reactor is designed with a thermal power output of 500 MWt. Modified CANDLE burn-up scheme allows the reactor to have long life operation by supplying only natural uranium as fuel cycle input. This scheme introducing discrete region, the fuel is initially put in region 1, after one cycle of 10 years of burn up it is shifted to region 2 and region 1 is filled by fresh natural uranium fuel. The reactor is designed for 100 years with 10 regions arranged axially. The results of neutronic calculation showed that the void coefficients ranged from -0.6695443 % at BOC to -0.5273626 % at EOC for 500 MWt reactor. The void coefficients of Pb-208 more negative than Pb-nat. The results showed that the reactors with Pb-208 coolant have better level of safety than Pb-nat.

  15. Tritium Burn-up Depth and Tritium Break-Even Time

    Institute of Scientific and Technical Information of China (English)

    LI Cheng-Yue; DENG Bai-Quan; HUANG Jin-Hua; YAN Jian-Cheng

    2006-01-01

    @@ Similarly to but quite different from the xenon poisoning effects resulting from fission-produced iodine during the restart-up process of a fission reactor, we introduce a completely new concept of the tritium burn-up depth and tritium break-even time in the fusion energy research area. To show what the least required amount of tritium storage is used to start up a fusion reactor and how long a time the fusion reactor needs to be operated for achieving the tritium break-even during the initial start-up phase due to the finite tritium breeding time that is dependent on the tritium breeder, specific structure of breeding zone, layout of coolant flow pipe, tritium recovery scheme, extraction process, the tritium retention of reactor components, unrecoverable tritium fraction in breeder, leakage to the inertial gas container, and the natural decay etc., we describe this new phenomenon and answer this problem by setting up and by solving a set of equations, which express a dynamic subsystem model of the tritium inventory evolution in a fusion experimental breeder (FEB). It is found that the tritium burn-up depth is 317g and the tritium break-even time is approximately 240 full power days for FEB designed detail configuration and it is also found that after one-year operation, the tritium storage reaches 1.18kg that is more than theleast required amount of tritium storage to start up three of FEB-like fusion reactors.

  16. Fuel burnup analysis for IRIS reactor using MCNPX and WIMS-D5 codes

    Science.gov (United States)

    Amin, E. A.; Bashter, I. I.; Hassan, Nabil M.; Mustafa, S. S.

    2017-02-01

    International Reactor Innovative and Secure (IRIS) reactor is a compact power reactor designed with especial features. It contains Integral Fuel Burnable Absorber (IFBA). The core is heterogeneous both axially and radially. This work provides the full core burn up analysis for IRIS reactor using MCNPX and WIMDS-D5 codes. Criticality calculations, radial and axial power distributions and nuclear peaking factor at the different stages of burnup were studied. Effective multiplication factor values for the core were estimated by coupling MCNPX code with WIMS-D5 code and compared with SAS2H/KENO-V code values at different stages of burnup. The two calculation codes show good agreement and correlation. The values of radial and axial powers for the full core were also compared with published results given by SAS2H/KENO-V code (at the beginning and end of reactor operation). The behavior of both radial and axial power distribution is quiet similar to the other data published by SAS2H/KENO-V code. The peaking factor values estimated in the present work are close to its values calculated by SAS2H/KENO-V code.

  17. Modeling of burnup express-estimation for UO{sub 2}-fuel

    Energy Technology Data Exchange (ETDEWEB)

    Likhanskii, Vladimir V.; Tokarev, Sergey A.; Vilkhivskaya, Olga V., E-mail: vilhivskaya_olga@mail.ru

    2017-03-15

    Highlights: • Proposed engineering model estimates fuel burnup by {sup 134}Cs/{sup 137}Cs activity ratio. • Buildup of cesium isotopes relies on changing neutron spectrum in the core cycle. • {sup 134}Cs/{sup 137}Cs activity ratios in FAs with Gd-doped fuel rods are analyzed. • Comparison of the model calculations with the NPPs spike measurements is presented. - Abstract: The paper presents the developed engineering model of cesium isotopes production as function of UO{sub 2}-fuel burnup and an assessment of their activity ratios. The model considers the evolution of linear power of gadolinium-doped fuel rods and fuel rods surrounding them in fuel assemblies with high enrichment fuel, harder neutron spectrum, and the changes in cross-sections of neutron reactions in thermal and epithermal energy areas. Parametrical dependences in the model are based on the fuel operation data for nuclear power plants and on the detailed neutronic-physical calculations of the core. Presented are the results of the model calculations for the {sup 134}Cs/{sup 137}Cs activity ratios in fuel taking into account the parameter of hardness of the neutron spectrum during the first irradiation cycle for fuel with enrichment ranging from 3.6 wt% in {sup 235}U.

  18. Comparison of neutron cross sections for selected fission products and isotopic composition analyses with burnup

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. H.; Gil, C. S.; Kim, J. D.; Jang, J. H.; Lee, Y. D. [KAERI, Taejon (Korea)

    2003-10-01

    The neutron absorption cross sections for 18 fission products evaluated within the framework of the KAERI-BNL international collaboration have been compared with the ENDF/B-VI release 7. Also, the influence of the new evaluations on isotopic compositions of the fission products as a function of burnup has been analyzed through the OECD/NEA burnup credit criticality benchmarks (Phase 1B) and the LWR/Pu recycling benchmarks. These calculations were performed by WIMSD-5B with the 69 group libraries prepared from three evaluated nuclear data libraries: ENDF/B-VI.7, ENDF/B-VI.8 including new evaluations in resonance region covering thermal region, and ENDF/B-VII expected including those in upper resonance region up to 20 MeV. For Xe-131, the composition calculated with ENDF/B-VI.8 shows maximum difference of 4.78% compared to ENDF/B-VI.7. However, the isotopic compositions of all fission products calculated with ENDF/B-VII shows no differences compared to ENDF/B-VI.7.

  19. Instant release fraction and matrix release of high burn-up UO2 spent nuclear fuel: Effect of high burn-up structure and leaching solution composition

    Science.gov (United States)

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

    2012-08-01

    Two weak points in Performance Assessment (PA) exercises regarding the alteration of Spent Nuclear Fuel (SNF) are the contribution of the so-called Instant Release Fraction (IRF) and the effect of High Burn-Up Structure (HBS). This manuscript focuses on the effect of HBS in matrix (long term) and instant release of a Pressurised Water Reactor (PWR) SNF irradiated in a commercial reactor with a mean Burn-Up (BU) of 60 GWd/tU. In order to study the HBS contribution, two samples from different radial positions have been prepared. One from the centre of the SNF, labelled CORE, and one from the periphery, enriched with HBS and labelled OUT. Static leaching experiments have been carried out with two synthetic leaching solutions: bicarbonate (BIC) and Bentonitic Granitic Groundwater (BGW), and in all cases under oxidising conditions. IRF values have been calculated from the determined Fraction of Inventory in Aqueous Phase (FIAP). In all studied cases, some radionuclides (RN): Rb, Sr and Cs, have shown higher release rates than uranium, especially at the beginning of the experiment, and have been considered as IRF. Redox sensitive RN like Mo and Tc have been found to dissolve slightly faster than uranium and further studies might be needed to confirm if they can also be considered part of the IRF. Most of the remaining studied RN, mainly actinides and lanthanides, have been found to dissolve congruently with the uranium matrix. Finally, Zr, Ru and Rh presented lower release rates than the matrix. Higher matrix release has been determined for CORE than for OUT samples showing that the formation of HBS might have a protective effect against the oxidative corrosion of the SNF. On the contrary, no significant differences have been observed between the two studied leaching solutions (BIC and BGW). Two different IRF contributions have been determined. One corresponding to the fraction of inventory segregated in the external open grain boundaries, directly available to water and

  20. Assessment of Fission Product Cross-Section Data for Burnup Credit Applications

    Energy Technology Data Exchange (ETDEWEB)

    Leal, Luiz C [ORNL; Derrien, Herve [ORNL; Dunn, Michael E [ORNL; Mueller, Don [ORNL

    2007-12-01

    Past efforts by the Department of Energy (DOE), the Electric Power Research Institute (EPRI), the Nuclear Regulatory Commission (NRC), and others have provided sufficient technical information to enable the NRC to issue regulatory guidance for implementation of pressurized-water reactor (PWR) burnup credit; however, consideration of only the reactivity change due to the major actinides is recommended in the guidance. Moreover, DOE, NRC, and EPRI have noted the need for additional scientific and technical data to justify expanding PWR burnup credit to include fission product (FP) nuclides and enable burnup credit implementation for boiling-water reactor (BWR) spent nuclear fuel (SNF). The criticality safety assessment needed for burnup credit applications will utilize computational analyses of packages containing SNF with FP nuclides. Over the years, significant efforts have been devoted to the nuclear data evaluation of major isotopes pertinent to reactor applications (i.e., uranium, plutonium, etc.); however, efforts to evaluate FP cross-section data in the resonance region have been less thorough relative to actinide data. In particular, resonance region cross-section measurements with corresponding R-matrix resonance analyses have not been performed for FP nuclides. Therefore, the objective of this work is to assess the status and performance of existing FP cross-section and cross-section uncertainty data in the resonance region for use in burnup credit analyses. Recommendations for new cross-section measurements and/or evaluations are made based on the data assessment. The assessment focuses on seven primary FP isotopes (103Rh, 133Cs, 143Nd, 149Sm, 151Sm, 152Sm, and 155Gd) that impact reactivity analyses of transportation packages and two FP isotopes (153Eu and 155Eu) that impact prediction of 155Gd concentrations. Much of the assessment work was completed in 2005, and the assessment focused on the latest FP cross-section evaluations available in the

  1. The new Polish nitriding and nitriding like processes in the modern technology

    Energy Technology Data Exchange (ETDEWEB)

    Has, Z.; Kula, P. [Technical Univ. of Lodz (Poland)

    1995-12-31

    Modern technological methods for making nitrided layers and low-friction combined layers have been described. The possibilities of structures and properties forming were analyzed as well as the area and examples of application were considered. Nitrided layers are applied in high loaded frictional couples, widely. They can be formed on steel or cast iron machine parts by the classic gas nitriding process or by modern numerous nitriding technologies.

  2. Phase identification of iron nitrides and iron oxy-nitrides with Mossbauer spectroscopy

    NARCIS (Netherlands)

    Borsa, DM; Boerma, DO

    2003-01-01

    The Mossbauer spectroscopy of all known Fe nitrides is the topic of this paper. Most of the data were accumulated during a study of the growth of the various Fe nitride phases using molecular beam epitaxy of Fe in the presence of a flux of atomic N, or by post-nitriding freshly grown Fe layers also

  3. Composite Reinforcement using Boron Nitride Nanotubes

    Science.gov (United States)

    2014-05-09

    Final 3. DATES COVERED (From - To) 11-Mar-2013 to 10-Mar-2014 4. TITLE AND SUBTITLE Composite Reinforcement using Boron Nitride Nanotubes...AVAILABILITY STATEMENT Approved for public release. 13. SUPPLEMENTARY NOTES 14. ABSTRACT Boron nitride nanotubes have been proposed as a...and titanium (Ti) metal clusters with boron nitride nanotubes (BNNT). First-principles density-functional theory plus dispersion (DFT-D) calculations

  4. Analysis of plasma-nitrided steels

    Science.gov (United States)

    Salik, J.; Ferrante, J.; Honecy, F.; Hoffman, R., Jr.

    1986-01-01

    The analysis of plasma nitrided steels can be divided to two main categories - structural and chemical. Structural analysis can provide information not only on the hardening mechanisms but also on the fundamental processes involved. Chemical analysis can be used to study the kinetics for the nitriding process and its mechanisms. In this paper preliminary results obtained by several techniques of both categories are presented and the applicability of those techniques to the analysis of plasma-nitrided steels is discussed.

  5. Ceramic processing of boron nitride insulators

    Energy Technology Data Exchange (ETDEWEB)

    Morgan, C. S.; McCulloch, R. W.

    1977-01-01

    Fuel pin simulators (FPS) are the prime elements of several test facilities at the Oak Ridge National Laboratory (ORNL). These experimental facilities are used to conduct out-of-reactor thermal-hydraulic and mechanical interaction safety tests for both light-water and breeder reactor programs. The FPS units simulate the geometry, heat flux profiles, and operational capabilities of a reactor core element under steady-state and transient conditions. They are subjected to temperatures as high as 1600/sup 0/C (2900/sup 0/F) and power levels as high as 57.5 kW/m (17.5 kW/ft) as well as severe thermal stresses during transient tests. The insulating material in the narrow annulus between the heating coil and the FPS sheath is subjected to very rigorous conditions. Accuracy of the reactor safety test information and validity of the test data depend on the heat flux uniformity under all test conditions and on the reliable operation of all fuel pin simulators and their internal thermocouples. Boron nitride (BN), because of its high degree of chemical inertness combined with its relatively unique properties of high thermal conductivity and low electrical conductivity, is the most suitable insulating material for FPS. The important BN properties, thermal conductivity and electrical resistance, are strongly influenced by crystallite orientation and by impurities. The article describes new BN powder processing techniques, which optimize these properties.

  6. Kinetic parameters study based on burn-up for improving the performance of research reactor equilibrium core

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2014-01-01

    Full Text Available In this study kinetic parameters, effective delayed neutron fraction and prompt neutron generation time have been investigated at different burn-up stages for research reactor's equilibrium core utilizing low enriched uranium high density fuel (U3Si2-Al fuel with 4.8 g/cm3 of uranium. Results have been compared with reference operating core of Pakistan research Reactor-1. It was observed that by increasing fuel burn-up, effective delayed neutron fraction is decreased while prompt neutron generation time is increased. However, over all ratio beff/L is decreased with increasing burn-up. Prompt neutron generation time L in the understudy core is lower than reference operating core of reactor at all burn-up steps due to hard spectrum. It is observed that beff is larger in the understudy core than reference operating core of due to smaller size. Calculations were performed with the help of computer codes WIMSD/4 and CITATION.

  7. 78 FR 67348 - Invitation for Public Comment on Draft Test Plan for the High Burnup Dry Storage Cask Research...

    Science.gov (United States)

    2013-11-12

    ...: U.S. Department of Energy, C/O Melissa Bates, 1955 Freemont Ave., MS 1235, Idaho Falls, ID 83415..., 1955 Fremont Ave., Attn: Melissa Bates, Idaho Falls, ID, between 8 a.m. and 3:30 p.m. MT, Monday.... Melissa Bates, Contracting Officers Representative, High Burnup Dry Storage Cask Research and...

  8. Conceptual design study on an upgraded future Monju core (2). Core concept with extended refueling interval and increased fuel burnup

    Energy Technology Data Exchange (ETDEWEB)

    Kinjo, Hidehito; Ishibashi, Jun-ichi; Nishi, Hiroshi [Japan Nuclear Cycle Development Inst., Tsuruga Head Office, International Cooperation and Technology Development Center, Tsuruga, Fukui (Japan); Kageyama, Takeshi [Nuclear Energy System Inc., Tokyo (Japan)

    2003-03-01

    A conceptual design study has been performed at the International Cooperation and Technology Development Center to investigate the feasibility of upgraded future Monju cores with extended refueling intervals of 365efpd/cycle and increased fuel burnup of 150 GWd/t. The goal of this study is to demonstrate the possible contribution of Monju to the improved economy and to efficient utilization, as one of the major facilities for fast neutron irradiation. Two design measures have been mainly taken to improve the core fuel burnup and reactivity control characteristics for the extended operating cycle length of 1 year: (1) The driver fuel pin specification with both increased pin diameter of 7.7mm and increased active core height of about 100cm has been chosen to reduce the burnup reactivity swing, (2) The absorber control rod specification has also been changed to enhance the control rod reactivity worth by increasing {sup 10}B-enrichment and absorber length, and to adequately secure the shutdown reactivity margin. The major core characteristics have been evaluated on the core power distribution, safety parameters such as sodium void reactivity and Doppler effect, thermal hydraulics and reactivity control characteristics. The results show that this core could achieve the targeted core performances of 1-year operating cycle as well as 150GWd/t discharged burnup, without causing any significant drawback on the core characteristics and safety aspects. The upgraded core concepts have, therefore, been confirmed as feasible. (author)

  9. Nucleation of iron nitrides during gaseous nitriding of iron; the effect of a preoxidation treatment

    DEFF Research Database (Denmark)

    Friehling, Peter B.; Poulsen, Finn Willy; Somers, Marcel A.J.

    2001-01-01

    grains. On prolonged nitriding, immediate nucleation at the surface of iron grains becomes possible. Calculated incubation times for the nucleation of gamma'-Fe4N1-x during nitriding are generally longer than those observed experimentally in the present work. The incubation time is reduced dramatically......The nucleation of iron nitrides during gaseous nitriding has been investigated using light microscopy and X-ray diffraction. Initially, the nucleation of gamma'-Fe4N1-x on a pure iron surface starts at grain boundaries meeting the surface, from where the nitride grains grow laterally into the iron...

  10. Applicability of the MCNP-ACAB system to inventory prediction in high-burnup fuels: sensitivity/uncertainty estimates

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Herranz, N.; Cabellos, O. [Madrid Polytechnic Univ., Dept. of Nuclear Engineering (Spain); Cabellos, O.; Sanz, J. [Madrid Polytechnic Univ., 2 Instituto de Fusion Nuclear (Spain); Sanz, J. [Univ. Nacional Educacion a Distancia, Dept. of Power Engineering, Madrid (Spain)

    2005-07-01

    We present a new code system which combines the Monte Carlo neutron transport code MCNP-4C and the inventory code ACAB as a suitable tool for high burnup calculations. Our main goal is to show that the system, by means of ACAB capabilities, enables us to assess the impact of neutron cross section uncertainties on the inventory and other inventory-related responses in high burnup applications. The potential impact of nuclear data uncertainties on some response parameters may be large, but only very few codes exist which can treat this effect. In fact, some of the most reported effective code systems in dealing with high burnup problems, such as CASMO-4, MCODE and MONTEBURNS, lack this capability. As first step, the potential of our system, ruling out the uncertainty capability, has been compared with that of those code systems, using a well referenced high burnup pin-cell benchmark exercise. It is proved that the inclusion of ACAB in the system allows to obtain results at least as reliable as those obtained using other inventory codes, such as ORIGEN2. Later on, the uncertainty analysis methodology implemented in ACAB, including both the sensitivity-uncertainty method and the uncertainty analysis by the Monte Carlo technique, is applied to this benchmark problem. We estimate the errors due to activation cross section uncertainties in the prediction of the isotopic content up to the high-burnup spent fuel regime. The most relevant uncertainties are remarked, and some of the most contributing cross sections to those uncertainties are identified. For instance, the most critical reaction for Am{sup 242m} is Am{sup 241}(n,{gamma}-m). At 100 MWd/kg, the cross-section uncertainty of this reaction induces an error of 6.63% on the Am{sup 242m} concentration.The uncertainties in the inventory of fission products reach up to 30%.

  11. Development, implementation, and verification of multicycle depletion perturbation theory for reactor burnup analysis

    Energy Technology Data Exchange (ETDEWEB)

    White, J.R.

    1980-08-01

    A generalized depletion perturbation formulation based on the quasi-static method for solving realistic multicycle reactor depletion problems is developed and implemented within the VENTURE/BURNER modular code system. The present development extends the original formulation derived by M.L. Williams to include nuclide discontinuities such as fuel shuffling and discharge. This theory is first described in detail with particular emphasis given to the similarity of the forward and adjoint quasi-static burnup equations. The specific algorithm and computational methods utilized to solve the adjoint problem within the newly developed DEPTH (Depletion Perturbation Theory) module are then briefly discussed. Finally, the main features and computational accuracy of this new method are illustrated through its application to several representative reactor depletion problems.

  12. Utilizing the burnup capability in MCNPX to perform depletion analysis of an MNSR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Boafo, Emmanuel [Ghana atomic Energy Commission, Accra (Ghana)

    2013-07-01

    The burnup capability in the MCNPX code was utilized to perform fuel depletion analysis of the MNSR LEU core by estimating the amount of fissile material (U-235) consumed as well as the amount of plutonium formed after the reactor core expected life. The decay heat removal rate for the MNSR after reactor shutdown was also investigated due to its significance to reactor safety. The results show that 0.568 % of U-235 was burnt up after 200 days of reactor operation while the amount of plutonium formed was not significant. The study also found that the decay heat decreased exponentially after reactor shutdown confirming that the decay heat will be removed from the system by natural circulation after shut down and hence safety of the reactor is assured.

  13. Development, implementation, and verification of multicycle depletion perturbation theory for reactor burnup analysis

    Energy Technology Data Exchange (ETDEWEB)

    White, J.R.

    1980-08-01

    A generalized depletion perturbation formulation based on the quasi-static method for solving realistic multicycle reactor depletion problems is developed and implemented within the VENTURE/BURNER modular code system. The present development extends the original formulation derived by M.L. Williams to include nuclide discontinuities such as fuel shuffling and discharge. This theory is first described in detail with particular emphasis given to the similarity of the forward and adjoint quasi-static burnup equations. The specific algorithm and computational methods utilized to solve the adjoint problem within the newly developed DEPTH (Depletion Perturbation Theory) module are then briefly discussed. Finally, the main features and computational accuracy of this new method are illustrated through its application to several representative reactor depletion problems.

  14. Burn-Up Determination by High Resolution Gamma Spectrometry: Axial and Diametral Scanning Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Forsyth, R.S.; Blackadder, W.H.; Ronqvist, N.

    1967-02-15

    In the gamma spectrometric determination of burn-up the use of a single fission product as a monitor of the specimen fission rate is subject to errors caused by activity saturation or, in certain cases, fission product migration. Results are presented of experiments in which all the resolvable gamma peaks in the fission product spectrum have been used to calculate the fission rate; these results form a pattern which reflect errors in the literature values of the gamma branching ratios, fission yields etc., and also represent a series of empirical correction factors. Axial and diametral scanning experiments on a long-irradiated low-enrichment fuel element are also described and demonstrate that it is possible to differentiate between fissions in U-235 and in Pu-239 respectively by means of the ratios of the Ru-106 activity to the activities of the other fission products.

  15. Mechanical Fatigue Testing of High-Burnup Fuel for Transportation Applications

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL; Wang, Hong [ORNL

    2015-05-01

    This report describes testing designed to determine the ability of high burnup (HBU) (>45 GWd/MTU) spent fuel to maintain its integrity under normal conditions of transportation. An innovative system, Cyclic Integrated Reversible-bending Fatigue Tester (CIRFT), has been developed at Oak Ridge National Laboratory (ORNL) to test and evaluate the mechanical behavior of spent nuclear fuel (SNF) under conditions relevant to storage and transportation. The CIRFT system is composed of a U-frame equipped with load cells for imposing the pure bending loads on the SNF rod test specimen and measuring the in-situ curvature of the fuel rod during bending using a set up with three linear variable differential transformers (LVDTs).

  16. Data Mining Techniques to Estimate Plutonium, Initial Enrichment, Burnup, and Cooling Time in Spent Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Trellue, Holly Renee [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fugate, Michael Lynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tobin, Stephen Joesph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-03-19

    The Next Generation Safeguards Initiative (NGSI), Office of Nonproliferation and Arms Control (NPAC), National Nuclear Security Administration (NNSA) of the U.S. Department of Energy (DOE) has sponsored a multi-laboratory, university, international partner collaboration to (1) detect replaced or missing pins from spent fuel assemblies (SFA) to confirm item integrity and deter diversion, (2) determine plutonium mass and related plutonium and uranium fissile mass parameters in SFAs, and (3) verify initial enrichment (IE), burnup (BU), and cooling time (CT) of facility declaration for SFAs. A wide variety of nondestructive assay (NDA) techniques were researched to achieve these goals [Veal, 2010 and Humphrey, 2012]. In addition, the project includes two related activities with facility-specific benefits: (1) determination of heat content and (2) determination of reactivity (multiplication). In this research, a subset of 11 integrated NDA techniques was researched using data mining solutions at Los Alamos National Laboratory (LANL) for their ability to achieve the above goals.

  17. Effects of microstructural constraints on the transport of fission products in uranium dioxide at low burnups

    Science.gov (United States)

    Lim, Harn Chyi; Rudman, Karin; Krishnan, Kapil; McDonald, Robert; Dickerson, Patricia; Gong, Bowen; Peralta, Pedro

    2016-08-01

    Diffusion of fission gases in UO2 is studied at low burnups, before bubble growth and coalescence along grain boundaries (GBs) become dominant, using a 3-D finite element model that incorporates actual UO2 microstructures. Grain boundary diffusivities are assigned based on crystallography with lattice and GB diffusion coupled with temperature to account for temperature gradients. Heterogeneity of GB properties and connectivity can induce regions where concentration is locally higher than without GB diffusion. These regions are produced by "bottlenecks" in the GB network because of lack of connectivity among high diffusivity GBs due to crystallographic constraints, and they can lead to localized swelling. Effective diffusivities were calculated assuming a uniform distribution of high diffusivity among GBs. Results indicate an increase over the bulk diffusivity with a clear grain size effect and that connectivity and properties of different GBs become important factors on the variability of fission product concentration at the microscale.

  18. Accident source terms for boiling water reactors with high burnup cores.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Powers, Dana Auburn; Leonard, Mark Thomas

    2007-11-01

    The primary objective of this report is to provide the technical basis for development of recommendations for updates to the NUREG-1465 Source Term for BWRs that will extend its applicability to accidents involving high burnup (HBU) cores. However, a secondary objective is to re-examine the fundamental characteristics of the prescription for fission product release to containment described by NUREG-1465. This secondary objective is motivated by an interest to understand the extent to which research into the release and behaviors of radionuclides under accident conditions has altered best-estimate calculations of the integral response of BWRs to severe core damage sequences and the resulting radiological source terms to containment. This report, therefore, documents specific results of fission product source term analyses that will form the basis for the HBU supplement to NUREG-1465. However, commentary is also provided on observed differences between the composite results of the source term calculations performed here and those reflected NUREG-1465 itself.

  19. Investigation of the Fundamental Constants Stability Based on the Reactor Oklo Burn-Up Analysis

    Science.gov (United States)

    Onegin, M. S.; Yudkevich, M. S.; Gomin, E. A.

    2012-12-01

    The burn-up of few samples of the natural Oklo reactor zones 3, 5 was calculated using the modern Monte Carlo code. We reconstructed the neutron spectrum in the core by means of the isotope ratios: 147Sm/148Sm and 176Lu/175Lu. These ratios unambiguously determine the water content and core temperature. The isotope ratio of the 149Sm in the sample calculated using this spectrum was compared with experimental one. The disagreement between these two values allows one to limit a possible shift of the low lying resonance of 149Sm. Then, these limits were converted to the limits for the change of the fine structure constant α. We have found out, that for the rate of α change, the inequality ěrt˙ {α }/α ěrt<= 5× 10-18 is fulfilled, which is one order higher than our previous limit.

  20. Investigation of the fundamental constants stability based on the reactor Oklo burn-up analysis

    CERN Document Server

    Onegin, M S

    2010-01-01

    The burn-up for SC56-1472 sample of the natural Oklo reactor zone 3 was calculated using the modern Monte Carlo codes. We reconstructed the neutron spectrum in the core by means of the isotope ratios: $^{147}$Sm/$^{148}$Sm and $^{176}$Lu/$^{175}$Lu. These ratios unambiguously determine the spectrum index and core temperature. The effective neutron absorption cross section of $^{149}$Sm calculated using this spectrum was compared with experimental one. The disagreement between these two values allows to limit a possible shift of the low laying resonance of $^{149}$Sm even more . Then, these limits were converted to the limits for the change of the fine structure constant $\\alpha$. We found that for the rate of $\\alpha$ change the inequality $|\\delta \\dot{\\alpha}/\\alpha| \\le 5\\cdot 10^{-18}$ is fulfilled, which is of the next higher order than our previous limit.

  1. Chemical forms of solid fission products in the irradiated uranium—plutonium mixed nitride fuel

    Science.gov (United States)

    Arai, Yasuo; Maeda, Atsushi; Shiozawa, Ken-ichi; Ohmichi, Toshihiko

    1994-06-01

    Chemical forms of solid fission products in the irradiated (U, Pu)N fuel were estimated by both thermodynamic equilibrium calculation and electron microprobe analysis on burnup simulated samples prepared by carbothermic reduction. Besides the MX type matrix phase dissolving zirconium, niobium, yttrium and rare earth elements, the existence of two kinds of inclusion was recognized. One is URu 3 type intermetallic compound constituted by uranium and platinum group elements. The other is an alloy containing molybdenum as a principal constituent. Furthermore, the swelling rate due to solid fission products precipitation was evaluated to be about 0.5% per %FIMA.

  2. Evaluation of fission product worth margins in PWR spent nuclear fuel burnup credit calculations.

    Energy Technology Data Exchange (ETDEWEB)

    Blomquist, R.N.; Finck, P.J.; Jammes, C.; Stenberg, C.G.

    1999-02-17

    Current criticality safety calculations for the transportation of irradiated LWR fuel make the very conservative assumption that the fuel is fresh. This results in a very substantial overprediction of the actual k{sub eff} of the transportation casks; in certain cases, this decreases the amount of spent fuel which can be loaded in a cask, and increases the cost of transporting the spent fuel to the repository. Accounting for the change of reactivity due to fuel depletion is usually referred to as ''burnup credit.'' The US DOE is currently funding a program aimed at establishing an actinide only burnup credit methodology (in this case, the calculated reactivity takes into account the buildup or depletion of a limited number of actinides). This work is undergoing NRC review. While this methodology is being validated on a significant experimental basis, it implicitly relies on additional margins: in particular, the absorption of neutrons by certain actinides and by all fission products is not taken into account. This provides an important additional margin and helps guarantee that the methodology is conservative provided these neglected absorption are known with reasonable accuracy. This report establishes the accuracy of fission product absorption rate calculations: (1) the analysis of European fission product worth experiments demonstrates that fission product cross-sections available in the US provide very good predictions of fission product worth; (2) this is confirmed by a direct comparison of European and US cross section evaluations; (3) accuracy of Spent Nuclear Fuel (SNF) fission product content predictions is established in a recent ORNL report where several SNF isotopic assays are analyzed; and (4) these data are then combined to establish in a conservative manner the fraction of the predicted total fission product absorption which can be guaranteed based on available experimental data.

  3. High burnup effects on fuel behaviour under accident conditions: the tests CABRI REP-Na

    Science.gov (United States)

    Schmitz, Franz; Papin, Joelle

    A large, performance based, knowledge and experience in the field of nuclear fuel behaviour is available for nominal operation conditions. The database is continuously completed and precursor assembly irradiations are performed for testing of new materials and innovative designs. This procedure produces data and arguments to extend licencing limits in the permanent research for economic competitiveness. A similar effort must be devoted to the establishment of a database for fuel behaviour under off-normal and accident conditions. In particular, special attention must be given to the so-called design-basis-accident (DBA) conditions. Safety criteria are formulated for these situations and must be respected without consideration of the occurrence probability and the risk associated to the accident situation. The introduction of MOX fuel into the cores of light water reactors and the steadily increasing goal burnup of the fuel call for research work, both experimental and analytical, in the field of fuel response to DBA conditions. In 1992, a significant programme step, CABRI REP-Na, has been launched by the French Nuclear Safety and Protection Institute (IPSN) in the field of the reactivity initiated accident (RIA). After performing the nine experiments of the initial test matrix it can be concluded that important new findings have been evidenced. High burnup clad corrosion and the associated degradation of the mechanical properties of the ZIRCALOY4 clad is one of the key phenomena of the fuel behaviour under accident conditions. Equally important is the evidence that transient, dynamic fission gas effects resulting from the close to adiabatic heating introduces a new explosive loading mechanism which may lead to clad rupture under RIA conditions, especially in the case of heterogeneous MOX fuel.

  4. Low temperature anodic bonding to silicon nitride

    DEFF Research Database (Denmark)

    Weichel, Steen; Reus, Roger De; Bouaidat, Salim;

    2000-01-01

    Low-temperature anodic bonding to stoichiometric silicon nitride surfaces has been performed in the temperature range from 3508C to 4008C. It is shown that the bonding is improved considerably if the nitride surfaces are either oxidized or exposed to an oxygen plasma prior to the bonding. Both bulk...

  5. Composite Reinforcement using Boron Nitride Nanotubes

    Science.gov (United States)

    2016-11-15

    ApprovedOMB No. 0704-0188 The public reporting burden for this collection of information is estimated to average 1 hour per response, including the time for...nitride nanotubes change with the presence of atomic oxygen were also carried out. 15.  SUBJECT TERMS Nanotubes, Boron Nitride, Composites, Theoretical

  6. PECVD silicon nitride diaphragms for condenser microphones

    NARCIS (Netherlands)

    Scheeper, P.R.; Voorthuyzen, J.A.; Bergveld, P.

    1991-01-01

    The application of plasma-enhanced chemical vapour deposited (PECVD) silicon nitride as a diaphragm material for condenser microphones has been investigated. By means of adjusting the SiH4/NH3 gas-flow composition, silicon-rich silicon nitride films have been obtained with a relatively low tensile s

  7. Method of preparation of uranium nitride

    Science.gov (United States)

    Kiplinger, Jaqueline Loetsch; Thomson, Robert Kenneth James

    2013-07-09

    Method for producing terminal uranium nitride complexes comprising providing a suitable starting material comprising uranium; oxidizing the starting material with a suitable oxidant to produce one or more uranium(IV)-azide complexes; and, sufficiently irradiating the uranium(IV)-azide complexes to produce the terminal uranium nitride complexes.

  8. Thermodynamics, kinetics and process control of nitriding

    DEFF Research Database (Denmark)

    Mittemeijer, Eric J.; Somers, Marcel A. J.

    1999-01-01

    , the nitriding result is determined largely by the kinetics of the process. The nitriding kinetics have been shown to be characterised by the occurring local near-equilibria and stationary states at surfaces and interfaces, and the diffusion coefficient of nitrogen in the various phases, for which new data have...

  9. Qualification of the B and W Mark B fuel assembly for high burnup. Third semi-annual progress report, July-December 1979

    Energy Technology Data Exchange (ETDEWEB)

    Coleman, T.A.

    1980-03-01

    Five Babcock and Wilcox-designed Mark B (15 x 15) pressurized water reactor fuel assemblies were irradiated to extended burnups in Duke Power Company's Oconee Unit 1 reactor. An assembly average burnup of 40,000 MWd/mtU, which is about 29% greater than previous discharge burnups at Oconee 1, was attained. The nondestructive examination of these five assemblies, which have been irradiated for four fuel cycles, was begun. Data obtained included fuel assembly and fuel dimensions, water channel spacings, fuel rod surface deposit samples, and holddown spring preload forces. Visual examination of the assemblies indicated that good fuel performance was maintained through four cycles of irradiation.

  10. Cathodic Cage Plasma Nitriding: An Innovative Technique

    Directory of Open Access Journals (Sweden)

    R. R. M. de Sousa

    2012-01-01

    Full Text Available Cylindrical samples of AISI 1020, AISI 316, and AISI 420 steels, with different heights, were simultaneously treated by a new technique of ionic nitriding, entitled cathodic cage plasma nitriding (CCPN, in order to evaluate the efficiency of this technique to produce nitrided layers with better properties compared with those obtained using conventional ionic nitriding technique. This method is able to eliminate the edge effect in the samples, promoting a better uniformity of temperature, and consequently, a smaller variation of the thickness/height relation can be obtained. The compound layers were characterized by X-ray diffraction, optical microscopy, and microhardness test profile. The results were compared with the properties of samples obtained with the conventional nitriding, for the three steel types. It was verified that samples treated by CCPN process presented, at the same temperature, a better uniformity in the thickness and absence of the edge effect.

  11. Hard carbon nitride and method for preparing same

    Science.gov (United States)

    Haller, E.E.; Cohen, M.L.; Hansen, W.L.

    1992-05-05

    Novel crystalline [alpha](silicon nitride-like)-carbon nitride and [beta](silicon nitride-like)-carbon nitride are formed by sputtering carbon in the presence of a nitrogen atmosphere onto a single crystal germanium or silicon, respectively, substrate. 1 figure.

  12. Molten-Salt-Based Growth of Group III Nitrides

    Science.gov (United States)

    Waldrip, Karen E.; Tsao, Jeffrey Y.; Kerley, Thomas M.

    2008-10-14

    A method for growing Group III nitride materials using a molten halide salt as a solvent to solubilize the Group-III ions and nitride ions that react to form the Group III nitride material. The concentration of at least one of the nitride ion or Group III cation is determined by electrochemical generation of the ions.

  13. Solvothermal synthesis: a new route for preparing nitrides

    CERN Document Server

    Demazeau, G; Denis, A; Largeteau, A

    2002-01-01

    Solvothermal synthesis appears to be an interesting route for preparing nitrides such as gallium nitride and aluminium nitride, using ammonia as solvent. A nitriding additive is used to perform the reaction and, in the case of gallium nitride, is encapsulated by melt gallium. The syntheses are performed in the temperature range 400-800 deg. C and in the pressure range 100-200 MPa. The synthesized powders are characterized by x-ray diffraction and scanning electron microscopy. Finely divided gallium nitride GaN and aluminium nitride AlN, both with wurtzite-type structure, can be obtained by this route.

  14. Electronic structure and magnetic properties of doped Al1- x Ti x N ( x = 0.03, 0.25) compositions based on cubic aluminum nitride from ab initio simulation data

    Science.gov (United States)

    Bannikov, V. V.; Beketov, A. R.; Baranov, M. V.; Elagin, A. A.; Kudyakova, V. S.; Shishkin, R. A.

    2016-05-01

    The phase stability, electronic structure, and magnetic properties of Al1- x Ti x N compositions based on the metastable aluminum nitride modification with the rock-salt structure at low ( x = 0.03) and high ( x = 0.25) concentrations of titanium in the system have been investigated using the results of ab initio band calculations. It has been shown that, at low values of x, the partial substitution is characterized by a positive enthalpy, which, however, changes sign with an increase in the titanium concentration. According to the results of the band structure calculations, the doped compositions have electronic conductivity. For x = 0.03, titanium impurity atoms have local magnetic moments (˜0.6 μB), and the electronic spectrum is characterized by a 100% spin polarization of near-Fermi states. Some of the specific features of the chemical bonding in Al1- x Ti x N cubic phases have been considered.

  15. Process for the production of metal nitride sintered bodies and resultant silicon nitride and aluminum nitride sintered bodies

    Science.gov (United States)

    Yajima, S.; Omori, M.; Hayashi, J.; Kayano, H.; Hamano, M.

    1983-01-01

    A process for the manufacture of metal nitride sintered bodies, in particular, a process in which a mixture of metal nitrite powders is shaped and heated together with a binding agent is described. Of the metal nitrides Si3N4 and AIN were used especially frequently because of their excellent properties at high temperatures. The goal is to produce a process for metal nitride sintered bodies with high strength, high corrosion resistance, thermal shock resistance, thermal shock resistance, and avoidance of previously known faults.

  16. Subsurface Aluminum Nitride Formation in Iron-Aluminum Alloys

    Science.gov (United States)

    Bott, June H.

    , which gradually decreased the internal precipitation zones with increasing aluminum content. In samples containing 8 wt.% aluminum, a thick continuous oxide scale formed and prevented nitrogen and oxygen penetration into the bulk of the sample, thus preventing the formation of any internal precipitates. The effect of modifying the heating rate in pure N2 atmospheres was examined. Samples were heated over the course of 1, 10, or 100 minutes. Faster heating rates increased the aluminum content in the oxide scale on all samples. Additionally, these rapid heating rate samples had either had lower internal precipitation depths or no internal precipitates. Experiments were conducted in N2--2.5% H2/H 2O mixtures with varying dew points to lower the oxygen potential of the reaction gas and prevent the formation of external iron oxide scales. In the 3 and 5 wt.% Al alloys, this produced an internal aluminum-rich oxide band which inhibited further internal precipitation. Samples treated in atmospheres to simulate the reheat furnace combustion atmosphere experienced dramatically increased external oxidation in addition to inward growth of the oxide scale and internal precipitation of oxides and nitrides within the metal. The most important scientific findings of this dissertation are the dramatic effect of heating rate on modifying the external scale of the alloys presented and the presence of continuous internal oxide bands in several samples throughout the study. Oxidation studies typically occur for longer times and in higher oxygen contents than the present results, so the influence of heating rate is either largely unnoticed or is overcome by oxide growth at long times. Oxide bands have been observed in literature, but few aluminum oxide bands have been seen before this study. vi.

  17. Modelling the evolution of composition-and stress-depth profiles in austenitic stainless steels during low-temperature nitriding

    DEFF Research Database (Denmark)

    Jespersen, Freja Nygaard; Hattel, Jesper Henri; Somers, Marcel A. J.

    2016-01-01

    that accompanies the dissolution of high nitrogen contents in expanded austenite. An intriguing phenomenon during low-temperature nitriding is that the residual stresses evoked by dissolution of nitrogen in the solid state, affect the thermodynamics and the diffusion kinetics of nitrogen dissolution......Nitriding of stainless steel causes a surface zone of expanded austenite, which improves the wear resistance of the stainless steel while preserving the stainless behaviour. During nitriding huge residual stresses are introduced in the treated zone, arising from the volume expansion....... In the present paper solid mechanics was combined with thermodynamics and diffusion kinetics to simulate the evolution of composition-depth and stress-depth profiles resulting from nitriding. The model takes into account a composition-dependent diffusion coefficient of nitrogen in expanded austenite, short range...

  18. Method of synthesizing cubic system boron nitride

    Energy Technology Data Exchange (ETDEWEB)

    Yuzu, S.; Sumiya, H.; Degawa, J.

    1987-10-13

    A method is described for synthetically growing cubic system boron nitride crystals by using boron nitride sources, solvents for dissolving the boron nitride sources, and seed crystals under conditions of ultra-high pressure and high temperature for maintaining the cubic system boron nitride stable. The method comprises the following steps: preparing a synthesizing vessel having at least two chambers, arrayed in order in the synthesizing vessel so as to be heated according to a temperature gradient; placing the solvents having different eutectic temperatures in each chamber with respect to the boron nitride sources according to the temperature gradient; placing the boron nitride source in contact with a portion of each of the solvents heated at a relatively higher temperature and placing at least a seed crystal in a portion of each of the solvents heated at a relatively lower temperature; and growing at least one cubic system boron nitride crystal in each of the solvents in the chambers by heating the synthesizing vessel for establishing the temperature gradient while maintaining conditions of ultra-high pressure and high temperature.

  19. Half metallic ferromagnetism in alkali metal nitrides MN (M = Rb, Cs): A first principles study

    Energy Technology Data Exchange (ETDEWEB)

    Murugan, A., E-mail: rrpalanichamy@gmail.com; Rajeswarapalanichamy, R., E-mail: rrpalanichamy@gmail.com; Santhosh, M., E-mail: rrpalanichamy@gmail.com; Sudhapriyanga, G., E-mail: rrpalanichamy@gmail.com [Department of Physics, N.M.S.S.V.N College, Madurai, Tamilnadu-625019 (India); Kanagaprabha, S. [Department of Physics, Kamaraj College, Tuticorin, Tamil Nadu-628003 (India)

    2014-04-24

    The structural, electronic and elastic properties of two alkali metal nitrides (MN: M= Rb, Cs) are investigated by the first principles calculations based on density functional theory using the Vienna ab-initio simulation package. At ambient pressure the two nitrides are stable in ferromagnetic state with CsCl structure. The calculated lattice parameters are in good agreement with the available results. The electronic structure reveals that these materials are half metallic in nature. A pressure-induced structural phase transition from CsCl to ZB phase is observed in RbN and CsN.

  20. Transparent polycrystalline cubic silicon nitride

    Science.gov (United States)

    Nishiyama, Norimasa; Ishikawa, Ryo; Ohfuji, Hiroaki; Marquardt, Hauke; Kurnosov, Alexander; Taniguchi, Takashi; Kim, Byung-Nam; Yoshida, Hidehiro; Masuno, Atsunobu; Bednarcik, Jozef; Kulik, Eleonora; Ikuhara, Yuichi; Wakai, Fumihiro; Irifune, Tetsuo

    2017-01-01

    Glasses and single crystals have traditionally been used as optical windows. Recently, there has been a high demand for harder and tougher optical windows that are able to endure severe conditions. Transparent polycrystalline ceramics can fulfill this demand because of their superior mechanical properties. It is known that polycrystalline ceramics with a spinel structure in compositions of MgAl2O4 and aluminum oxynitride (γ-AlON) show high optical transparency. Here we report the synthesis of the hardest transparent spinel ceramic, i.e. polycrystalline cubic silicon nitride (c-Si3N4). This material shows an intrinsic optical transparency over a wide range of wavelengths below its band-gap energy (258 nm) and is categorized as one of the third hardest materials next to diamond and cubic boron nitride (cBN). Since the high temperature metastability of c-Si3N4 in air is superior to those of diamond and cBN, the transparent c-Si3N4 ceramic can potentially be used as a window under extremely severe conditions. PMID:28303948

  1. Friction Characteristics of Nitrided Layers on AISI 430 Ferritic Stainless Steel Obtained by Various Nitriding Processes

    Directory of Open Access Journals (Sweden)

    Hakan AYDIN

    2013-03-01

    Full Text Available The influence of plasma, gas and salt-bath nitriding techniques on the friction coefficient of AISI 430 ferritic stainless steel was studied in this paper. Samples were plasma nitrided in 80 % N2 + 20 % H2 atmosphere at 450 °C and 520 °C for 8 h at a pressure of 2 mbar, gas nitrided in NH3 and CO2 atmosphere at 570 °C for 13 h and salt-bath nitrided in a cyanide-cyanate salt-bath at 570 °C for 1.5 h. Characterisation of nitrided layers on the ferritic stainless steel was carried out by means of microstructure, microhardness, surface roughness and friction coefficient measurements. Friction characteristics of the nitrided layers on the 430 steel were investigated using a ball-on-disc friction-wear tester with a WC-Co ball as the counter-body under dry sliding conditions. Analysis of wear tracks was carried out by scanning electron microscopy. Maximum hardness and maximum case depth were achieved on the plasma nitrided sample at 520 ºC for 8 h. The plasma and salt-bath nitriding techniques significantly decreased the average surface roughness of the 430 ferritic stainless steel. The friction test results showed that the salt-bath nitrided layer had better friction-reducing ability than the other nitrided layers under dry sliding conditions. Furthermore, the friction characteristic of the plasma nitrided layer at 520 ºC was better than that of the plasma nitrided layer at 450 °C.DOI: http://dx.doi.org/10.5755/j01.ms.19.1.3819

  2. Evaluation technology for burnup and generated amount of plutonium by measurement of xenon isotopic ratio in dissolver off-gas at reprocessing facility (Joint research)

    OpenAIRE

    岡野 正紀; 久野 剛彦; 高橋 一朗; 白水 秀知; Charlton, W. S.; Wells, C. A.; Hemberger, P. H.; 山田 敬二; 酒井 敏雄

    2006-01-01

    The amount of Pu in the spent fuel was evaluated from Xe isotopic ratio in off-gas in reprocessing facility, is related to burnup. Six batches of dissolver off-gas at spent fuel dissolution process were sampled from the main stack in Tokai Reprocessing Plant during BWR fuel reprocessing campaign. Xenon isotopic ratio was determined with GC/MS. Burnup and generated amount of Pu were evaluated with Noble Gas Environmental Monitoring Application code (NOVA), developed by Los Alamos National Labo...

  3. Burnup determination of a fuel element concerning different cooling times; Seguimiento del quemado de un elemento combustible, para diferentes tiempos de enfriamento

    Energy Technology Data Exchange (ETDEWEB)

    Henriquez, C.; Navarro, G.; Pereda, C.; Mutis, O. [Comision Chilena de Energia Nuclear, Santiago (Chile). Dept. de Aplicaciones Nucleares. Unidad de Reactores; Terremoto, Luis A.A.; Zeituni, Carlos A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear

    2002-07-01

    In this work we report a complete set of measurements and some relevant results regarding the burnup process of a fuel element containing low enriched nuclear fuel. This fuel element was fabricated at the Plant of Fuel Elements of the Chilean Nuclear Energy Commission (CCHEN). Measurements were carried out using gamma-ray spectroscopy and the absolute burnup of the fuel element was determined. (author)

  4. Nanoporous Carbon Nitride: A High Efficient Filter for Seawater Desalination

    CERN Document Server

    Li, Weifeng; Zhou, Hongcai; Zhang, Xiaoming; Zhao, Mingwen

    2015-01-01

    The low efficiency of commercially-used reverse osmosis (RO) membranes has been the main obstacle in seawater desalination application. Here, we report the auspicious performance, through molecular dynamics simulations, of a seawater desalination filter based on the recently-synthesized graphene-like carbon nitride (g-C2N) [Nat. Commun., 2015, 6, 6486]. Taking advantage of the inherent nanopores and excellent mechanical properties of g-C2N filter, highly efficient seawater desalination can be achieved by modulating the nanopores under tensile strain. The water permeability can be improved by two orders of magnitude compared to RO membranes, which offers a promising approach to the global water shortage solution.

  5. Compressibility and thermal expansion of cubic silicon nitride

    DEFF Research Database (Denmark)

    Jiang, Jianzhong; Lindelov, H.; Gerward, Leif

    2002-01-01

    The compressibility and thermal expansion of the cubic silicon nitride (c-Si3N4) phase have been investigated by performing in situ x-ray powder-diffraction measurements using synchrotron radiation, complemented with computer simulations by means of first-principles calculations. The bulk...... compressibility of the c-Si3N4 phase originates from the average of both Si-N tetrahedral and octahedral compressibilities where the octahedral polyhedra are less compressible than the tetrahedral ones. The origin of the unit cell expansion is revealed to be due to the increase of the octahedral Si-N and N-N bond...

  6. Conceptual Design study of Small Long-life Gas Cooled Fast Reactor With Modified CANDLE Burn-up Scheme

    Science.gov (United States)

    Nur Asiah, A.; Su'ud, Zaki; Ferhat, A.; Sekimoto, H.

    2010-06-01

    In this paper, conceptual design study of Small Long-life Gas Cooled Fast Reactors with Natural Uranium as Fuel Cycle Input has been performed. In this study Gas Cooled Fast Reactor is slightly modified by employing modified CANDLE burn-up scheme so that it can use Natural Uranium as fuel cycle input. Due to their hard spectrum, GCFR in this study showed very good performance in converting U-238 to plutonium in order to maintain the operation condition requirement of long-life reactors. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. With such condition we got an optimal design of 325 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input. The average discharge burn-up is about 290 GWd/ton HM.

  7. Residual Stress Induced by Nitriding and Nitrocarburizing

    DEFF Research Database (Denmark)

    Somers, Marcel A.J.

    2005-01-01

    The present chapter is devoted to the various mechanisms involved in the buildup and relief of residual stress in nitrided and nitrocarburized cases. The work presented is an overview of model studies on iron and iron-based alloys. Subdivision is made between the compound (or white) layer......, developing at the surfce and consisting of iron-based (carbo)nitrides, and the diffusion zone underneath, consisting of iron and alloying element nitrides dispersed in af ferritic matrix. Microstructural features are related directly to the origins of stress buildup and stres relief....

  8. Plasma Nitriding of Low Alloy Sintered Steels

    Institute of Scientific and Technical Information of China (English)

    Shiva Mansoorzadeh; Fakhreddin Ashrafizadeh; Xiao-Ying Li; Tom Bell

    2004-01-01

    Fe-3Cr-0.5Mo-0.3C and Fe-3Cr-1.4Mn-0.5Mo-0.367C sintered alloys were plasma nitrided at different temperatures. Characterization was performed by microhardness measurement, optical microscopy, SEM and XRD. Both materials had similar nitriding case properties. 1.4% manganese did not change the as-sintered microstructure considerably.It was observed that monophase compound layer, γ, formed with increasing temperature. Compound layer thickness increased with increasing temperature while nitriding depth increased up to a level and then decreased. Core softening was more pronounced at higher temperature owing to cementite coarsening.

  9. Residual Stress Induced by Nitriding and Nitrocarburizing

    DEFF Research Database (Denmark)

    Somers, Marcel A.J.

    2005-01-01

    The present chapter is devoted to the various mechanisms involved in the buildup and relief of residual stress in nitrided and nitrocarburized cases. The work presented is an overview of model studies on iron and iron-based alloys. Subdivision is made between the compound (or white) layer......, developing at the surfce and consisting of iron-based (carbo)nitrides, and the diffusion zone underneath, consisting of iron and alloying element nitrides dispersed in af ferritic matrix. Microstructural features are related directly to the origins of stress buildup and stres relief....

  10. Atomic Resolution Microscopy of Nitrides in Steel

    DEFF Research Database (Denmark)

    Danielsen, Hilmar Kjartansson

    2014-01-01

    MN and CrMN type nitride precipitates in 12%Cr steels have been investigated using atomic resolution microscopy. The MN type nitrides were observed to transform into CrMN both by composition and crystallography as Cr diffuses from the matrix into the MN precipitates. Thus a change from one precip...... layer between the crystalline nitride and ferrite matrix. Usually precipitates are described as having (semi) coherent or incoherent interfaces, but in this case it is more energetically favourable to create an amorphous layer instead of the incoherent interface....

  11. Synthesis of ternary nitrides by mechanochemical alloying

    DEFF Research Database (Denmark)

    Jacobsen, C.J.H.; Zhu, J.J.; Lindelov, H.;

    2002-01-01

    Ternary metal nitrides ( of general formula MxM'N-y(z)) attract considerable interest because of their special mechanical, electrical, magnetic, and catalytic properties. Usually they are prepared by ammonolysis of ternary oxides (MxM'O-y(m)) at elevated temperatures. We show that ternary...... nitrides by mechanochemical alloying of a binary transition metal nitride (MxN) with an elemental transition metal. In this way, we have been able to prepare Fe3Mo3N and Co3Mo3N by ball-milling of Mo2N with Fe and Co, respectively. The transformation sequence from the starting materials ( the binary...

  12. Supramolecular intermediates in the synthesis of polymeric carbon nitride from melamine cyanurate

    Energy Technology Data Exchange (ETDEWEB)

    Dante, Roberto C., E-mail: rcdante@yahoo.com [Facultad de Mecánica, Escuela Politécnica Nacional (EPN), Ladrón de Guevara E11-253, Quito (Ecuador); Sánchez-Arévalo, Francisco M. [Instituto de Investigaciones en Materiales, Universidad Nacional Autónoma de Mexico, Apdo. Postal 70-360, Cd. Universitaria, Mexico D.F. 04510 (Mexico); Chamorro-Posada, Pedro [Dpto. de Teoría de la Señal y Comunicaciones e IT, Universidad de Valladolid, ETSI Telecomunicación, Paseo Belén 15, 47011 Valladolid (Spain); Vázquez-Cabo, José [Dpto. de Teoría de la Señal y Comunicaciones, Universidad de Vigo, ETSI Telecomunicación, Lagoas Marcosende s/n, Vigo (Spain); Huerta, Lazaro [Instituto de Investigaciones en Materiales, Universidad Nacional Autónoma de Mexico, Apdo. Postal 70-360, Cd. Universitaria, Mexico D.F. 04510 (Mexico); Lartundo-Rojas, Luis [Centro de Nanociencias y Micro y Nanotecnologías—IPN, Luis Enrique Erro s/n, U. Prof. Adolfo López Mateos, 07738 Ciudad de Mexico, Distrito Federal (Mexico); Santoyo-Salazar, Jaime [Departamento de Física, Centro de Investigación y de Estudios Avanzados del Instituto Politécnico Nacional, CINVESTAV-IPN, Apdo. Postal 14-740, Mexico D.F. 07360 (Mexico); and others

    2015-03-15

    The adduct of melamine and cyanuric acid (MCA) was used in past research to produce polymeric carbon nitride and precursors. The reaction yield was considerably incremented by the addition of sulfuric acid. The polymeric carbon nitride formation occurs around 450 °C at temperatures above the sublimation of the adduct components, which occurs around 400 °C. In this report the effect of sulfuric acid on MCA was investigated. It was found that the MCA rosette supramolecular channel structures behave as a solid solvent able to host small molecules, such as sulfuric acid, inside these channels and interact with them. Therefore, the sulfuric acid effect was found to be close to that of a solute that causes a temperature increment of the “solvent sublimation” enough to allowing the formation of polymeric carbon nitride to occur. Sulfate ions are presumably hosted in the rosette channels of MCA as shown by simulations. - Graphical abstract: The blend of melamine cyanurate and sulfuric acid behaves like a solution so that melamine cyanurate decomposition is shifted to temperatures high enough to react and form polymeric carbon nitride. - Highlights: • The adduct of melamine and cyanuric acid behaves as a solid solvent. • The blend of sulfuric acid and melamine cyanurate behaves like a solution. • Melamine cyanurate decomposition is shifted to higher temperatures by sulfuric acid. • The formation of polymeric carbon nitride occurs for these higher temperatures.

  13. Study on behavior of plasma nitrided 316L in PEMFC working conditions

    Energy Technology Data Exchange (ETDEWEB)

    Tian, Rujin [Institute of Materials and Technology, Dalian Maritime University, Linghailu No.1, Ganjingzi District, Dalian, Lianoing 116026 (China); College of Materials and Engineering, Dalian Jiaotong University, Dalian 116028 (China); Sun, Juncai [Institute of Materials and Technology, Dalian Maritime University, Linghailu No.1, Ganjingzi District, Dalian, Lianoing 116026 (China); Wang, Jianli [Department of Basic Science, Changchun University of Technology, Changchun 130012 (China)

    2008-12-15

    Stainless steel bipolar plates for the polymer electrolyte membrane fuel cell (PEMFC) offer many advantages over conventional machined graphite and graphite-composites. However, the interfacial ohmic loss between the metallic bipolar plate and membrane electrode assembly due to corrosion decreases the overall power output of PEMFC. A lower temperature (at 370 C) plasma nitriding was applied to modify the surface of stainless steel 316L bipolar plates. The results of electrochemical measurements show that corrosion resistance of the plasma nitrided 316L is improved in simulated PEMFC anode/cathode environments purged with H{sub 2}/air at 70 C. The surface conductivity of the nitrided layer is better than that of the air-formed oxide film. The interfacial contact resistance (ICR) between the passive film and carbon paper increases very little after potentiostatic polarization for 4 h, which indicates potential for good stability of this material in highly corrosive fuel cell environments. (author)

  14. Plasma nitriding of AISI 52100 ball bearing steel and effect of heat treatment on nitrided layer

    Indian Academy of Sciences (India)

    Ravindra Kumar; J Alphonsa; Ram Prakash; K S Boob; J Ghanshyam; P A Rayjada; P M Raole; S Mukherjee

    2011-02-01

    In this paper an effort has been made to plasma nitride the ball bearing steel AISI 52100. The difficulty with this specific steel is that its tempering temperature (∼170–200°C) is much lower than the standard processing temperature (∼460–580°C) needed for the plasma nitriding treatment. To understand the mechanism, effect of heat treatment on the nitrided layer steel is investigated. Experiments are performed on three different types of ball bearing races i.e. annealed, quenched and quench-tempered samples. Different gas compositions and process temperatures are maintained while nitriding these samples. In the quenched and quench-tempered samples, the surface hardness has decreased after plasma nitriding process. Plasma nitriding of annealed sample with argon and nitrogen gas mixture gives higher hardness in comparison to the hydrogen–nitrogen gas mixture. It is reported that the later heat treatment of the plasma nitrided annealed sample has shown improvement in the hardness of this steel. X-ray diffraction analysis shows that the dominant phases in the plasma nitrided annealed sample are (Fe2−3N) and (Fe4N), whereas in the plasma nitrided annealed sample with later heat treatment only -Fe peak occurs.

  15. Temperature and burnup correlated fuel-cladding chemical interaction in U-10ZR metallic fuel

    Science.gov (United States)

    Carmack, William J.

    Metallic fuels are proposed for use in advanced sodium cooled fast reactors and provide a number of advantages over other fuel types considering their fabricability, performance, recyclability, and safety. Resistance to cladding "breach" and subsequent release of fission products and fuel constituents to the nuclear power plant primary coolant system is a key performance parameter for a nuclear fuel system. In metallic fuel, FCCI weakens the cladding, especially at high power-high temperature operation, contributing to fuel pin breach. Empirical relationships for FCCI have been developed from a large body of data collected from in-pile (EBR-II) and out-of-pile experiments [1]. However, these relationships are unreliable in predicting FCCI outside the range of EBR-II experimental data. This dissertation examines new FCCI data extracted from the MFF-series of prototypic length metallic fuel irradiations performed in the Fast Flux Test Facility (FFTF). The fuel in these assemblies operated a temperature and burnup conditions similar to that in EBR-II but with axial fuel height three times longer than EBR-II experiments. Comparing FCCI formation data from FFTF and EBR-II provides new insight into FCCI formation kinetics. A model is developed combining both production and diffusion of lanthanides to the fuel-cladding interface and subsequent reaction with the cladding. The model allows these phenomena to be influenced by fuel burnup (lanthanide concentrations) and operating temperature. Parameters in the model are adjusted to reproduce measured FCCI layer thicknesses from EBR-II and FFTF. The model predicts that, under appropriate conditions, rate of FCCI formation can be controlled by either fission product transport or by the reaction rate of the interaction species at the fuel-cladding interface. This dissertation will help forward the design of metallic fuel systems for advanced sodium cooled fast reactors by allowing the prediction of FCCI layer formation in full

  16. Current applications of actinide-only burn-up credit within the Cogema group and R and D programme to take fission products into account

    Energy Technology Data Exchange (ETDEWEB)

    Toubon, H. [Cogema, 78 - Saint Quentin en Yvelines (France); Guillou, E. [Cogema Etablissement de la Hague, D/SQ/SMT, 50 - Beaumont Hague (France); Cousinou, P. [CEA Fontenay aux Roses, Inst. de Protection et de Surete Nucleaire, 92 (France); Barbry, F. [CEA Valduc, Inst. de Protection et de Surete Nucleaire, 21 - Is sur Tille (France); Grouiller, J.P.; Bignan, G. [CEA Cadarache, 13 - Saint Paul lez Durance (France)

    2001-07-01

    Burn-up credit can be defined as making allowance for absorbent radioactive isotopes in criticality studies, in order to optimise safety margins and avoid over-engineering of nuclear facilities. As far as the COGEMA Group is concerned, the three fields in which burn-up credit proves to be an advantage are the transport of spent fuel assemblies, their interim storage in spent fuel pools and reprocessing. In the case of transport, burn-up credit means that cask size do not need to be altered, despite an increase in the initial enrichment of the fuel assemblies. Burn-up credit also makes it possible to offer new cask designs with higher capacity. Burn-up credit means that fuel assemblies with a higher initial enrichment can be put into interim storage in existing facilities and opens the way to the possibility of more compact ones. As far as reprocessing is concerned, burn-up credit makes it possible to keep up current production rates, despite an increase in the initial enrichment of the fuel assemblies being reprocessed. In collaboration with the French Atomic Energy Commission and the Institute for Nuclear Safety and Protection, the COGEMA Group is participating in an extensive experimental programme and working to qualify criticality and fuel depletion computer codes. The research programme currently underway should mean that by 2003, allowance will be made for fission products in criticality safety analysis.

  17. A feasibility study to determine cooling time and burnup of ATR fuel using a nondestructive technique and three types of gamma-ray detectors

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, J.; Aryaeinejad, R.; Nigg, D.W. [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID 83415 (United States)

    2011-07-01

    The goal of this work was to perform a feasibility study and establish measurement techniques to determine the burnup of the Advanced Test Reactor (ATR) fuels at the Idaho National Laboratory (INL). Three different detectors of high purity germanium (HPGe), lanthanum bromide (LaBr{sub 3}), and high pressure xenon (HPXe) in two detection system configurations of below and above the water pool were used in this study. The last two detectors were used for the first time in fuel burnup measurements. The results showed that a better quality spectra can be achieved with the above the water pool configuration. Both short and long cooling time fuels were investigated in order to determine which measurement technique, absolute or fission product ratio, is better suited in each scenario and also to establish what type of detector should be used in each case for the best burnup measurement. The burnup and cooling time calibrations were established using experimental absolute activities or isotopic ratios and ORIGEN burnup calculations. A method was developed to do burnup and cooling time calibrations using fission isotopes activities without the need to know the exact geometry. (authors)

  18. Quantification of the computational accuracy of code systems on the burn-up credit using experimental re-calculations; Quantifizierung der Rechengenauigkeit von Codesystemen zum Abbrandkredit durch Experimentnachrechnungen

    Energy Technology Data Exchange (ETDEWEB)

    Behler, Matthias; Hannstein, Volker; Kilger, Robert; Moser, Franz-Eberhard; Pfeiffer, Arndt; Stuke, Maik

    2014-06-15

    In order to account for the reactivity-reducing effect of burn-up in the criticality safety analysis for systems with irradiated nuclear fuel (''burnup credit''), numerical methods to determine the enrichment and burnup dependent nuclide inventory (''burnup code'') and its resulting multiplication factor k{sub eff} (''criticality code'') are applied. To allow for reliable conclusions, for both calculation systems the systematic deviations of the calculation results from the respective true values, the bias and its uncertainty, are being quantified by calculation and analysis of a sufficient number of suitable experiments. This quantification is specific for the application case under scope and is also called validation. GRS has developed a methodology to validate a calculation system for the application of burnup credit in the criticality safety analysis for irradiated fuel assemblies from pressurized water reactors. This methodology was demonstrated by applying the GRS home-built KENOREST burnup code and the criticality calculation sequence CSAS5 from SCALE code package. It comprises a bounding approach and alternatively a stochastic, which both have been exemplarily demonstrated by use of a generic spent fuel pool rack and a generic dry storage cask, respectively. Based on publicly available post irradiation examination and criticality experiments, currently the isotopes of uranium and plutonium elements can be regarded for.

  19. Recent view to the results of pulse tests in the IGR reactor with high burn-up fuel

    Energy Technology Data Exchange (ETDEWEB)

    Asmolov, V.; Yegorova, L. [Russian Research Centre, Moscow (Russian Federation)

    1996-03-01

    Testing of 43 fuel elements (13 fuel elements with high burn-up fuel, 10 fuel elements with preirradiated cladding and fresh fuel, and 20 non-irradiated fuel elements) was carried out in the IGR pulse reactor with a half width of the reactor power pulse of about 0.7 sec. Tests were conducted in capsules with no coolant flow and with standard initial conditions in the capsule of 20{degrees}C and 0.2 MPa. Two types of coolant were used: water and air. One purpose of the test program was to determine the thresholds and mechanisms of fuel rod failure under RIA conditions for VVER fuel rods over their entire exposure range, from zero to high burn-up. These failure thresholds are often used in safety analyses. The tests and analyses were designed to reveal the influence on fuel rod failure of (1) the mechanical properties of the cladding, (2) the pellet-to-cladding gap, (3) fuel burn-up, (4) fuel-to-coolant heat transfer, and other parameters. The resulting data base can also be used for validation of computer codes used for analyzing fuel rod behavior. Three types of test specimens were used in the tests, and diagrams of these specimens are shown in Fig. 1. {open_quotes}Type-C{close_quotes} specimens were re-fabricated from commercial fuel rods of the VVER-1000 type that had been subjected to many power cycles of operation in the Novovoronezh Nuclear Power Plant (NV NPP). {open_quotes}Type-D{close_quotes} specimens were fabricated from the same commercial fuel rods used above, but the high burn-up oxide fuel was removed from the cladding and was replaced with fresh oxide fuel pellets. {open_quotes}Type-D{close_quotes} specimens thus provided a means of separating the effects of the cladding and the oxide fuel pellets and were used to examine cladding effects only.

  20. In Comparative Analysis for Fuel Burnup of Fuel Assembly Designs for the 300 kW Small Medical Reactor

    Science.gov (United States)

    Sambuu, Odmaa; Nanzad, Norov

    2009-03-01

    A 300 kW small medical reactor was designed to be used for boron neutron capture therapy (BNCT) at KAIST in 1996 [1]. In this paper, analysis for the core life cycle of the original design of the BNCT facility and modifications of the fuel assembly configuration and enrichment to get a proper life cycle were performed and a criticality, neutron flux distribution and fuel burnup calculations were carried out.

  1. Characterization of the non-uniqueness of used nuclear fuel burnup signatures through a Mesh-Adaptive Direct Search

    Energy Technology Data Exchange (ETDEWEB)

    Skutnik, Steven E., E-mail: sskutnik@utk.edu; Davis, David R.

    2016-05-01

    The use of passive gamma and neutron signatures from fission indicators is a common means of estimating used fuel burnup, enrichment, and cooling time. However, while characteristic fission product signatures such as {sup 134}Cs, {sup 137}Cs, {sup 154}Eu, and others are generally reliable estimators for used fuel burnup within the context where the assembly initial enrichment and the discharge time are known, in the absence of initial enrichment and/or cooling time information (such as when applying NDA measurements in a safeguards/verification context), these fission product indicators no longer yield a unique solution for assembly enrichment, burnup, and cooling time after discharge. Through the use of a new Mesh-Adaptive Direct Search (MADS) algorithm, it is possible to directly probe the shape of this “degeneracy space” characteristic of individual nuclides (and combinations thereof), both as a function of constrained parameters (such as the assembly irradiation history) and unconstrained parameters (e.g., the cooling time before measurement and the measurement precision for particular indicator nuclides). In doing so, this affords the identification of potential means of narrowing the uncertainty space of potential assembly enrichment, burnup, and cooling time combinations, thereby bounding estimates of assembly plutonium content. In particular, combinations of gamma-emitting nuclides with distinct half-lives (e.g., {sup 134}Cs with {sup 137}Cs and {sup 154}Eu) in conjunction with gross neutron counting (via {sup 244}Cm) are able to reasonably constrain the degeneracy space of possible solutions to a space small enough to perform useful discrimination and verification of fuel assemblies based on their irradiation history.

  2. OECD/NEA burnup credit criticality benchmarks phase IIIA: Criticality calculations of BWR spent fuel assemblies in storage and transport

    Energy Technology Data Exchange (ETDEWEB)

    Okuno, Hiroshi; Naito, Yoshitaka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ando, Yoshihira [Toshiba Corp., Kawasaki, Kanagawa (Japan)

    2000-09-01

    The report describes the final results of Phase IIIA Benchmarks conducted by the Burnup Credit Criticality Calculation Working Group under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD/NEA). The benchmarks are intended to confirm the predictive capability of the current computer code and data library combinations for the neutron multiplication factor (k{sub eff}) of a layer of irradiated BWR fuel assembly array model. In total 22 benchmark problems are proposed for calculations of k{sub eff}. The effects of following parameters are investigated: cooling time, inclusion/exclusion of FP nuclides and axial burnup profile, and inclusion of axial profile of void fraction or constant void fractions during burnup. Axial profiles of fractional fission rates are further requested for five cases out of the 22 problems. Twenty-one sets of results are presented, contributed by 17 institutes from 9 countries. The relative dispersion of k{sub eff} values calculated by the participants from the mean value is almost within the band of {+-}1%{delta}k/k. The deviations from the averaged calculated fission rate profiles are found to be within {+-}5% for most cases. (author)

  3. The calculational VVER burnup Credit Benchmark No.3 results with the ENDF/B-VI rev.5 (1999)

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez Gual, Maritza [Centro de Tecnologia Nuclear, La Habana (Cuba). E-mail: mrgual@ctn.isctn.edu.cu

    2000-07-01

    The purpose of this papers to present the results of CB3 phase of the VVER calculational benchmark with the recent evaluated nuclear data library ENDF/B-VI Rev.5 (1999). This results are compared with the obtained from the other participants in the calculations (Czech Republic, Finland, Hungary, Slovaquia, Spain and the United Kingdom). The phase (CB3) of the VVER calculation benchmark is similar to the Phase II-A of the OECD/NEA/INSC BUC Working Group benchmark for PWR. The cases without burnup profile (BP) were performed with the WIMS/D-4 code. The rest of the cases have been carried with DOTIII discrete ordinates code. The neutron library used was the ENDF/B-VI rev. 5 (1999). The WIMS/D-4 (69 groups) is used to collapse cross sections from the ENDF/B-VI Rev. 5 (1999) to 36 groups working library for 2-D calculations. This work also comprises the results of CB1 (obtained with ENDF/B-VI rev. 5 (1999), too) and CB3 for cases with Burnup of 30 MWd/TU and cooling time of 1 and 5 years and for case with Burnup of 40 MWd/TU and cooling time of 1 year. (author)

  4. High Frequency Acoustic Microscopy for the Determination of Porosity and Young's Modulus in High Burnup Uranium Dioxide Nuclear Fuel

    Science.gov (United States)

    Marchetti, Mara; Laux, Didier; Cappia, Fabiola; Laurie, M.; Van Uffelen, P.; Rondinella, V. V.; Wiss, T.; Despaux, G.

    2016-06-01

    During irradiation UO2 nuclear fuel experiences the development of a non-uniform distribution of porosity which contributes to establish varying mechanical properties along the radius of the pellet. Radial variations of both porosity and elastic properties in high burnup UO2 pellet can be investigated via high frequency acoustic microscopy. For this purpose ultrasound waves are generated by a piezoelectric transducer and focused on the sample, after having travelled through a coupling liquid. The elastic properties of the material are related to the velocity of the generated Rayleigh surface wave (VR). A UO2 pellet with a burnup of 67 GWd/tU was characterized using the acoustic microscope installed in the hot cells of the JRC-ITU at a 90 MHz frequency, with methanol as coupling liquid. VR was measured at different radial positions. A good agreement was found, when comparing the porosity values obtained via acoustic microscopy with those determined using SEM image analysis, especially in the areas close to the centre. In addition, Young's modulus was calculated and its radial profile was correlated to the corresponding burnup profile and to the hardness radial profile data obtained by Vickers micro-indentation.

  5. Surface modification of titanium by plasma nitriding

    Directory of Open Access Journals (Sweden)

    Kapczinski Myriam Pereira

    2003-01-01

    Full Text Available A systematic investigation was undertaken on commercially pure titanium submitted to plasma nitriding. Thirteen different sets of operational parameters (nitriding time, sample temperature and plasma atmosphere were used. Surface analyses were performed using X-ray diffraction, nuclear reaction and scanning electron microscopy. Wear tests were done with stainless steel Gracey scaler, sonic apparatus and pin-on-disc machine. The obtained results indicate that the tribological performance can be improved for samples treated with the following conditions: nitriding time of 3 h; plasma atmosphere consisting of 80%N2+20%H2 or 20%N2+80%H2; sample temperature during nitriding of 600 or 800 degreesC.

  6. Titanium nitride nanoparticles for therapeutic applications

    DEFF Research Database (Denmark)

    Guler, Urcan; Kildishev, Alexander V.; Boltasseva, Alexandra;

    2014-01-01

    Titanium nitride nanoparticles exhibit plasmonic resonances in the biological transparency window where high absorption efficiencies can be obtained with small dimensions. Both lithographic and colloidal samples are examined from the perspective of nanoparticle thermal therapy. © 2014 OSA....

  7. Materials synthesis: Two-dimensional gallium nitride

    Science.gov (United States)

    Koratkar, Nikhil A.

    2016-11-01

    Graphene is used as a capping sheet to synthesize 2D gallium nitride by means of migration-enhanced encapsulation growth. This technique may allow the stabilization of 2D materials that are not amenable to synthesis by traditional methods.

  8. Dissolution of bulk specimens of silicon nitride

    Science.gov (United States)

    Davis, W. F.; Merkle, E. J.

    1981-01-01

    An accurate chemical characterization of silicon nitride has become important in connection with current efforts to incorporate components of this material into advanced heat engines. However, there are problems concerning a chemical analysis of bulk silicon nitride. Current analytical methods require the pulverization of bulk specimens. A pulverization procedure making use of grinding media, on the other hand, will introduce contaminants. A description is given of a dissolution procedure which overcomes these difficulties. It has been found that up to at least 0.6 g solid pieces of various samples of hot pressed and reaction bonded silicon nitride can be decomposed in a mixture of 3 mL hydrofluoric acid and 1 mL nitric acid overnight at 150 C in a Parr bomb. High-purity silicon nitride is completely soluble in nitric acid after treatment in the bomb. Following decomposition, silicon and hydrofluoric acid are volatilized and insoluble fluorides are converted to a soluble form.

  9. The effect of dissolved hydrogen on the dissolution of {sup 233}U doped UO{sub 2}(s) high burn-up spent fuel and MOX fuel

    Energy Technology Data Exchange (ETDEWEB)

    Carbol, P. [Inst. for Transuranium Elements, Karlsruhe (Germany); Spahiu, K. (ed.) [and others

    2005-03-01

    In this report the results of the experimental work carried out in a large EU-research project (SFS, 2001-2004) on spent fuel stability in the presence of various amounts of near field hydrogen are presented. Studies of the dissolution of {sup 233}U doped UO{sub 2}(s) simulating 'old' spent fuel were carried out as static leaching tests, autoclave tests with various hydrogen concentrations and electrochemical tests. The results of the leaching behaviour of a high burn-up spent fuel pellet in 5 M NaCl solutions in the presence of 3.2 bar H{sub 2} pressure and of MOX fuel in dilute synthetic groundwater under 53 bar H{sub 2} pressure are also presented. In all the experimental studies carried out in this project, a considerable effect of hydrogen in the dissolution rates of radioactive materials was observed. The experimental results obtained in this project with a-doped UO{sub 2}, high burn-up spent fuel and MOX fuel together with literature data give a reliable background to use fractional alteration/dissolution rates for spent fuel of the order of 10{sup -6}/yr - 10{sup -8}/yr with a recommended value of 4x10{sup -7}/yr for dissolved hydrogen concentrations above 10{sup -3} M and Fe(II) concentrations typical for European repository concepts. Finally, based on a review of the experimental data and available literature data, potential mechanisms of the hydrogen effect are also discussed. The work reported in this document was performed as part of the Project SFS of the European Commission 5th Framework Programme under contract no FIKW-CT-2001-20192 SFS. It represents the deliverable D10 of the experimental work package 'Key experiments using a-doped UO{sub 2} and real spent fuel', coordinated by SKB with the participation of ITU, FZK-INE, ENRESA, CIEMAT, ARMINES-SUBATECH and SKB.

  10. Fuel Burnup and Fuel Pool Shielding Analysis for Bushehr Nuclear Reactor VVER-1000

    Science.gov (United States)

    Hadad, Kamal; Ayobian, Navid

    Bushehr Nuclear power plant (BNPP) is currently under construction. The VVER-1000 reactor will be loaded with 126 tons of about 4% enriched fuel having 3-years life cycle. The spent fuel (SF) will be transferred into the spent fuel pool (SPF), where it stays for 8 years before being transferred to Russia. The SPF plays a crucial role during 8 years when the SP resides in there. This paper investigates the shielding of this structure as it is designed to shield the SF radiation. In this study, the SF isotope inventory, for different cycles and with different burnups, was calculated using WIMS/4D transport code. Using MCNP4C nuclear code, the intensity of γ rays was obtained in different layers of SFP shields. These layers include the water above fuel assemblies (FA) in pool, concrete wall of the pool and water laid above transferring fuels. Results show that γ rays leakage from the shield in the mentioned layers are in agreement with the plant's PSAR data. Finally we analyzed an accident were the water height above the FA in the pool drops to 47 cm. In this case it was observed that exposure dose above pool, 10 and 30 days from the accident, are still high and in the levels of 1000 and 758 R/hr.

  11. Development and verification of fuel burn-up calculation model in a reduced reactor geometry

    Energy Technology Data Exchange (ETDEWEB)

    Sembiring, Tagor Malem [Center for Reactor Technology and Nuclear Safety (PTKRN), National Nuclear Energy Agency (BATAN), Kawasan PUSPIPTEK Gd. No. 80, Serpong, Tangerang 15310 (Indonesia)], E-mail: tagorms@batan.go.id; Liem, Peng Hong [Research Laboratory for Nuclear Reactor (RLNR), Tokyo Institute of Technology (Tokyo Tech), O-okayama, Meguro-ku, Tokyo 152-8550 (Japan)

    2008-02-15

    A fuel burn-up model in a reduced reactor geometry (2-D) is successfully developed and implemented in the Batan in-core fuel management code, Batan-FUEL. Considering the bank mode operation of the control rods, several interpolation functions are investigated which best approximate the 3-D fuel assembly radial power distributions across the core as function of insertion depth of the control rods. Concerning the applicability of the interpolation functions, it can be concluded that the optimal coefficients of the interpolation functions are not very sensitive to the core configuration and core or fuel composition in RSG GAS (MPR-30) reactor. Consequently, once the optimal interpolation function and its coefficients are derived then they can be used for 2-D routine operational in-core fuel management without repeating the expensive 3-D neutron diffusion calculations. At the selected fuel elements (at H-9 and G-6 core grid positions), the discrepancy of the FECFs (fuel element channel power peaking factors) between the 2-D and 3-D models are within the range of 3.637 x 10{sup -4}, 3.241 x 10{sup -4} and 7.556 x 10{sup -4} for the oxide, silicide cores with 250 g {sup 235}U/FE and the silicide core with 300 g {sup 235}U/FE, respectively.

  12. Propagation of Nuclear Data Uncertainties for ELECTRA Burn-up Calculations

    Science.gov (United States)

    Sjöstrand, H.; Alhassan, E.; Duan, J.; Gustavsson, C.; Koning, A. J.; Pomp, S.; Rochman, D.; Österlund, M.

    2014-04-01

    The European Lead-Cooled Training Reactor (ELECTRA) has been proposed as a training reactor for fast systems within the Swedish nuclear program. It is a low-power fast reactor cooled by pure liquid lead. In this work, we propagate the uncertainties in 239Pu transport data to uncertainties in the fuel inventory of ELECTRA during the reactor lifetime using the Total Monte Carlo approach (TMC). Within the TENDL project, nuclear models input parameters were randomized within their uncertainties and 740 239Pu nuclear data libraries were generated. These libraries are used as inputs to reactor codes, in our case SERPENT, to perform uncertainty analysis of nuclear reactor inventory during burn-up. The uncertainty in the inventory determines uncertainties in: the long-term radio-toxicity, the decay heat, the evolution of reactivity parameters, gas pressure and volatile fission product content. In this work, a methodology called fast TMC is utilized, which reduces the overall calculation time. The uncertainty of some minor actinides were observed to be rather large and therefore their impact on multiple recycling should be investigated further. It was also found that, criticality benchmarks can be used to reduce inventory uncertainties due to nuclear data. Further studies are needed to include fission yield uncertainties, more isotopes, and a larger set of benchmarks.

  13. Fuel burnup analysis of the TRIGA Mark II Reactor at the University of Pavia

    CERN Document Server

    Chiesa, Davide; Pozzi, Stefano; Previtali, Ezio; Sisti, Monica; Alloni, Daniele; Magrotti, Giovanni; Manera, Sergio; Prata, Michele; Salvini, Andrea; Cammi, Antonio; Zanetti, Matteo; Sartori, Alberto

    2015-01-01

    A time evolution model was developed to study fuel burnup for the TRIGA Mark II reactor at the University of Pavia. The results were used to predict the effects of a complete core reconfiguration and the accuracy of this prediction was tested experimentally. We used the Monte Carlo code MCNP5 to reproduce system neutronics in different operating conditions and to analyse neutron fluxes in the reactor core. The software that took care of time evolution, completely designed in-house, used the neutron fluxes obtained by MCNP5 to evaluate fuel consumption. This software was developed specifically to keep into account some features that differentiate experimental reactors from power ones, such as the daily ON/OFF cycle and the long fuel lifetime. These effects can not be neglected to properly account for neutron poison accumulation. We evaluated the effect of 48 years of reactor operation and predicted a possible new configuration for the reactor core: the objective was to remove some of the fuel elements from the...

  14. First steps towards a validation of the new burnup and depletion code TNT

    Energy Technology Data Exchange (ETDEWEB)

    Herber, S.C.; Allelein, H.J. [RWTH Aachen (Germany). Inst. for Reactor Safety and Reactor Technology; Research Center Juelich (Germany). Inst. for Energy and Climate Research - Nuclear Waste Disposal and Reactor Safety (IEK-6); Friege, N. [RWTH Aachen (Germany). Inst. for Reactor Safety and Reactor Technology; Kasselmann, S. [Research Center Juelich (Germany). Inst. for Energy and Climate Research - Nuclear Waste Disposal and Reactor Safety (IEK-6)

    2012-11-01

    In the frame of the fusion of the core design calculation capabilities, represented by V.S.O.P., and the accident calculation capabilities, represented by MGT(-3D), the successor of the TINTE code, difficulties were observed in defining an interface between a program backbone and the ORIGEN code respectively the ORIGENJUEL code. The estimation of the effort of refactoring the ORIGEN code or to write a new burnup code from scratch, led to the decision that it would be more efficient writing a new code, which could benefit from existing programming and software engineering tools from the computer code side and which can use the latest knowledge of nuclear reactions, e.g. consider all documented reaction channels. Therefore a new code with an object-oriented approach was developed at IEK-6. Object-oriented programming is currently state of the art and provides mostly an improved extensibility and maintainability. The new code was named TNT which stands for Topological Nuclide Transformation, since the code makes use of the real topology of the nuclear reactions. Here we want to present some first validation results from code to code benchmarks with the codes ORIGEN V2.2 and FISPACT2005 and whenever possible analytical results also used for the comparison. The 2 reference codes were chosen due to their high reputation in the field of fission reactor analysis (ORIGEN) and fusion facilities (FISPACT). (orig.)

  15. Fuel failure and fission gas release in high burnup PWR fuels under RIA conditions

    Science.gov (United States)

    Fuketa, Toyoshi; Sasajima, Hideo; Mori, Yukihide; Ishijima, Kiyomi

    1997-09-01

    To study the fuel behavior and to evaluate the fuel enthalpy threshold of fuel rod failure under reactivity initiated accident (RIA) conditions, a series of experiments using pulse irradiation capability of the Nuclear Safety Research Reactor (NSRR) has been performed. During the experiments with 50 MWd/kg U PWR fuel rods (HBO test series; an acronym for high burnup fuels irradiated in Ohi unit 1 reactor), significant cladding failure occurred. The energy deposition level at the instant of the fuel failure in the test is 60 cal/g fuel, and is considerably lower than those expected and pre-evaluated. The result suggests that mechanical interaction between the fuel pellets and the cladding tube with decreased integrity due to hydrogen embrittlement causes fuel failure at the low energy deposition level. After the pulse irradiation, the fuel pellets were found as fragmented debris in the coolant water, and most of these were finely fragmented. This paper describes several key observations in the NSRR experiments, which include cladding failure at the lower enthalpy level, possible post-failure events and large fission gas release.

  16. Cladding stress during extended storage of high burnup spent nuclear fuel

    Science.gov (United States)

    Raynaud, Patrick A. C.; Einziger, Robert E.

    2015-09-01

    In an effort to assess the potential for low temperature creep and delayed hydride cracking failures in high burnup spent fuel cladding during extended dry storage, the U.S. NRC analytical fuel performance tools were used to predict cladding stress during a 300 year dry storage period for UO2 fuel burned up to 65 GWd/MTU. Fuel swelling correlations were developed and used along with decay gas production and release fractions to produce circumferential average cladding stress predictions with the FRAPCON-3.5 fuel performance code. The resulting stresses did not result in cladding creep failures. The maximum creep strains accumulated were on the order of 0.54-1.04%, but creep failures are not expected below at least 2% strain. The potential for delayed hydride cracking was assessed by calculating the critical flaw size required to trigger this failure mechanism. The critical flaw size far exceeded any realistic flaw expected in spent fuel at end of reactor life.

  17. SEM Characterization of the High Burn-up Microstructure of U-7Mo Alloy

    Energy Technology Data Exchange (ETDEWEB)

    Dennis D. Keiser, Jr.; Jan-Fong Jue; Jian Gan; Brandon Miller; Adam Robinson; Pavel Medvedev; James Madden; Dan Wachs; M. Teague

    2014-04-01

    During irradiation, the microstructure of U-7Mo evolves until at a fission density near 5x1021 f/cm3 a high-burnup microstructure exists that is very different than what was observed at lower fission densities. This microstructure is dominated by randomly distributed, relatively large, homogeneous fission gas bubbles. The bubble superlattice has collapsed in many microstructural regions, and the fuel grain sizes, in many areas, become sub-micron in diameter with both amorphous fuel and crystalline fuel present. Solid fission product precipitates can be found inside the fission gas bubbles. To generate more information about the characteristics of the high-fission density microstructure, three samples irradiated in the RERTR-7 experiment have been characterized using a scanning electron microscope equipped with a focused ion beam. The FIB was used to generate samples for SEM imaging and to perform 3D reconstruction of the microstructure, which can be used to look for evidence of possible fission gas bubble interlinkage.

  18. Irradiation characteristics examination technology development of irradiated nuclear material and high burn-up fuels

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Kwon Pyo; Choo, Y. S.; Oh, Y. W. [and others

    2002-12-01

    The research and development for the first year of the project are performed through specialization of researchers, information from aborad and international cooperation, securement of advanced nuclear technology, development and installation of test equipment, application of external man-power, establishment of advanced test techniques, and certified test method. 1. Absolute efficiency measurement examination technology development of gamma scanning system 2. Sample preparation technology development of SEM and EPMA for micro-structural observation and chemical composition analysis 3. Irradiated high burn-up nuclear fuel transportation and test for PWR 4. Development of hot cell examination techniques and equipment 5. Acquirement of KOLAS system. In addition to the project, the following activities are carried out as follows; - PIE of Hanaro fuel(KH99H-001) - PIE of U-Mo advanced nuclear fuel irradiated at Hanaro - PIE of Hi-MET advanced nuclear fuel irradiated at Hanaro - PIE of DUPIC project - Hot cell examination of Hanaro irradiated capsule - Leaching test of PWR fuels - Surveillance test of PWR vessels - Mechanical test of CANDU pressure tubes.

  19. Propagation of nuclear data uncertainties for ELECTRA burn-up calculations

    CERN Document Server

    ostrand, H; Duan, J; Gustavsson, C; Koning, A; Pomp, S; Rochman, D; Osterlund, M

    2013-01-01

    The European Lead-Cooled Training Reactor (ELECTRA) has been proposed as a training reactor for fast systems within the Swedish nuclear program. It is a low-power fast reactor cooled by pure liquid lead. In this work, we propagate the uncertainties in Pu-239 transport data to uncertainties in the fuel inventory of ELECTRA during the reactor life using the Total Monte Carlo approach (TMC). Within the TENDL project the nuclear models input parameters were randomized within their uncertainties and 740 Pu-239 nuclear data libraries were generated. These libraries are used as inputs to reactor codes, in our case SERPENT, to perform uncertainty analysis of nuclear reactor inventory during burn-up. The uncertainty in the inventory determines uncertainties in: the long-term radio-toxicity, the decay heat, the evolution of reactivity parameters, gas pressure and volatile fission product content. In this work, a methodology called fast TMC is utilized, which reduces the overall calculation time. The uncertainty in the ...

  20. TEM Characterization of High Burn-up Microstructure of U-7Mo Alloy

    Energy Technology Data Exchange (ETDEWEB)

    Jian Gan; Brandon Miller; Dennis Keiser; Adam Robinson; James Madden; Pavel Medvedev; Daniel Wachs

    2014-04-01

    As an essential part of global nuclear non-proliferation effort, the RERTR program is developing low enriched U-Mo fuels (< 20% U-235) for use in research and test reactors that currently employ highly enriched uranium fuels. One type of fuel being developed is a dispersion fuel plate comprised of U-7Mo particles dispersed in Al alloy matrix. Recent TEM characterizations of the ATR irradiated U-7Mo dispersion fuel plates include the samples with a local fission densities of 4.5, 5.2, 5.6 and 6.3 E+21 fissions/cm3 and irradiation temperatures of 101-136?C. The development of the irradiated microstructure of the U-7Mo fuel particles consists of fission gas bubble superlattice, large gas bubbles, solid fission product precipitates and their association to the large gas bubbles, grain subdivision to tens or hundreds of nanometer size, collapse of bubble superlattice, and amorphisation. This presentation will describe the observed microstructures specifically focusing on the U-7Mo fuel particles. The impact of the observed microstructure on the fuel performance and the comparison of the relevant features with that of the high burn-up UO2 fuels will be discussed.

  1. Reticulated porous silicon nitride-based ceramics

    OpenAIRE

    Mazzocchi, Mauro; Medri, Valentina; Guicciardi, Stefano

    2012-01-01

    The interest towards the production of porous silicon nitride originates from the unique combination of light weight, of mechanical and physical properties typical of this class of ceramics that make them attractive for many engineering applications. Although pores are generally believed to deteriorate the mechanical properties of ceramics (the strength of porous ceramics decreases exponentially with an increase of porosity), the recent literature reports that porous silicon nitride can exhib...

  2. The Nitrogen-Nitride Anode.

    Energy Technology Data Exchange (ETDEWEB)

    Delnick, Frank M.

    2014-10-01

    Nitrogen gas N 2 can be reduced to nitride N -3 in molten LiCl-KCl eutectic salt electrolyte. However, the direct oxidation of N -3 back to N 2 is kinetically slow and only occurs at high overvoltage. The overvoltage for N -3 oxidation can be eliminated by coordinating the N -3 with BN to form the dinitridoborate (BN 2 -3 ) anion which forms a 1-D conjugated linear inorganic polymer with -Li-N-B-N- repeating units. This polymer precipitates out of solution as Li 3 BN 2 which becomes a metallic conductor upon delithiation. Li 3 BN 2 is oxidized to Li + + N 2 + BN at about the N 2 /N -3 redox potential with very little overvoltage. In this report we evaluate the N 2 /N -3 redox couple as a battery anode for energy storage.

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

    Science.gov (United States)

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

    2013-09-01

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

  4. Modelling of the layer evolution during nitriding processes

    Energy Technology Data Exchange (ETDEWEB)

    Figueroa, U.; Oseguera, J.; Schabes, P. [CEM, Atizapan (Mexico)

    1995-12-31

    The evolution of concomitant layers of nitrides is presented. The layer formation is experimentally achieved through two processes: Nitriding with a weakly ionized plasma and nitrogen post-discharge nitriding. The nitriding processes were performed on samples of pure iron and carbon steel. Nitriding temperatures were close but different from the eutectoid transformation point temperature. The experimental layer growth pattern is compared with a model of mass transfer, in which interface mass balance is considered. In the model the authors have considered the formation of one and two compact nitride layers. For short time of treatment, it is shown that a parabolic profile does not satisfactorily describe the layer growth.

  5. Electrochemical Solution Growth of Magnetic Nitrides

    Energy Technology Data Exchange (ETDEWEB)

    Monson, Todd C. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Pearce, Charles [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-10-01

    Magnetic nitrides, if manufactured in bulk form, would provide designers of transformers and inductors with a new class of better performing and affordable soft magnetic materials. According to experimental results from thin films and/or theoretical calculations, magnetic nitrides would have magnetic moments well in excess of current state of the art soft magnets. Furthermore, magnetic nitrides would have higher resistivities than current transformer core materials and therefore not require the use of laminates of inactive material to limit eddy current losses. However, almost all of the magnetic nitrides have been elusive except in difficult to reproduce thin films or as inclusions in another material. Now, through its ability to reduce atmospheric nitrogen, the electrochemical solution growth (ESG) technique can bring highly sought after (and previously inaccessible) new magnetic nitrides into existence in bulk form. This method utilizes a molten salt as a solvent to solubilize metal cations and nitrogen ions produced electrochemically and form nitrogen compounds. Unlike other growth methods, the scalable ESG process can sustain high growth rates (~mm/hr) even under reasonable operating conditions (atmospheric pressure and 500 °C). Ultimately, this translates into a high throughput, low cost, manufacturing process. The ESG process has already been used successfully to grow high quality GaN. Below, the experimental results of an exploratory express LDRD project to access the viability of the ESG technique to grow magnetic nitrides will be presented.

  6. Multi-objective optimization of steel nitriding

    Directory of Open Access Journals (Sweden)

    P. Cavaliere

    2016-03-01

    Full Text Available Steel nitriding is a thermo-chemical process largely employed in the machine components production to solve mainly wear and fatigue damage in materials. The process is strongly influenced by many different variables such as steel composition, nitrogen potential (range 0.8–35, temperature (range 350–1200 °C, time (range 2–180 hours. In the present study, the influence of such parameters affecting the nitriding layers' thickness, hardness, composition and residual stress was evaluated. The aim was to streamline the process by numerical–experimental analysis allowing to define the optimal conditions for the success of the process. The optimization software that was used is modeFRONTIER (Esteco, through which was defined a set of input parameters (steel composition, nitrogen potential, nitriding time, etc. evaluated on the basis of an optimization algorithm carefully chosen for the multi-objective analysis. The mechanical and microstructural results belonging to the nitriding process, performed with different processing conditions for various steels, are presented. The data were employed to obtain the analytical equations describing nitriding behavior as a function of nitriding parameters and steel composition. The obtained model was validated through control designs and optimized by taking into account physical and processing conditions.

  7. Molecular Dynamics Simulations of Hypervelocity Impacts

    Science.gov (United States)

    Owens, Eli T.; Bachlechner, Martina E.

    2007-03-01

    Outer space silicon solar cells are exposed to impacts with micro meteors that can destroy the surface leading to device failure. A protective coating of silicon nitride will protect against such failure. Large-scale molecular dynamics simulations are used to study how silicon/silicon nitride fails due to hypervelocity impacts. Three impactors made of silicon nitride are studied. Their cross-sectional areas, relative to the target, are as follows: the same as the target, half of the target, and a quarter of the target. Impactor speeds from 5 to 11 km/second yield several modes of failure, such as deformation of the target by the impactor and delimitation of the silicon nitride from the silicon at the interface. These simulations will give a much clearer picture of how solar cells composed of a silicon/silicon nitride interface will respond to impacts in outer space. This will ultimately lead to improved devices with longer life spans.

  8. Physics-based multiscale modeling of III-nitride light emitters

    OpenAIRE

    Zhou, Xiangyu

    2016-01-01

    The application of computer simulations to scientific and engineering problems has evolved to an established phase over the last decades. In the field of semiconductor device physics, Technology CAD (TCAD) has been regarded as an indispensable tool for the interpretation and prediction of device behavior. More specifically, TCAD modeling and simulation of nanostructured III-nitride light emitters still have challenging problems and is currently a topic under active research. This thesis devot...

  9. Direct access to macroporous chromium nitride and chromium titanium nitride with inverse opal structure.

    Science.gov (United States)

    Zhao, Weitian; DiSalvo, Francis J

    2015-03-21

    We report a facile synthesis of single-phase, nanocrystalline macroporous chromium nitride and chromium titanium nitride with an inverse opal morphology. The material is characterized using XRD, SEM, HR-TEM/STEM, TGA and XPS. Interconversion of macroporous CrN to Cr2O3 and back to CrN while retaining the inverse opal morphology is also demonstrated.

  10. Junctions between a boron nitride nanotube and a boron nitride sheet.

    Science.gov (United States)

    Baowan, Duangkamon; Cox, Barry J; Hill, James M

    2008-02-20

    For future nanoelectromechanical signalling devices, it is vital to understand how to connect various nanostructures. Since boron nitride nanostructures are believed to be good electronic materials, in this paper we elucidate the classification of defect geometries for combining boron nitride structures. Specifically, we determine possible joining structures between a boron nitride nanotube and a flat sheet of hexagonal boron nitride. Firstly, we determine the appropriate defect configurations on which the tube can be connected, given that the energetically favourable rings for boron nitride structures are rings with an even number of sides. A new formula E = 6+2J relating the number of edges E and the number of joining positions J is established for each defect, and the number of possible distinct defects is related to the so-called necklace and bracelet problems of combinatorial theory. Two least squares approaches, which involve variation in bond length and variation in bond angle, are employed to determine the perpendicular connection of both zigzag and armchair boron nitride nanotubes with a boron nitride sheet. Here, three boron nitride tubes, which are (3, 3), (6, 0) and (9, 0) tubes, are joined with the sheet, and Euler's theorem is used to verify geometrically that the connected structures are sound, and their relationship with the bonded potential energy function approach is discussed. For zigzag tubes (n,0), it is proved that such connections investigated here are possible only for n divisible by 3.

  11. Structural analysis of nitride layer formed on uranium metal by glow plasma surface nitriding

    Energy Technology Data Exchange (ETDEWEB)

    Liu Kezhao, E-mail: liukz@hotmail.com [State Key Laboratory of Silicon Materials and Department of Materials Science and Engineering, Zhejiang University, Hangzhou 310027 (China); Science and Technology on Surface Physics and Chemistry Laboratory, P.O. Box 718-35, Mianyang 621907 (China); Bin Ren [Science and Technology on Surface Physics and Chemistry Laboratory, P.O. Box 718-35, Mianyang 621907 (China); Xiao Hong [China Academy of Engineering Physics, P.O. Box 919-71, Mianyang 621907 (China); Long Zhong [State Key Laboratory of Silicon Materials and Department of Materials Science and Engineering, Zhejiang University, Hangzhou 310027 (China); Hong Zhanglian, E-mail: hong_zhanglian@zju.edu.cn [State Key Laboratory of Silicon Materials and Department of Materials Science and Engineering, Zhejiang University, Hangzhou 310027 (China); Yang Hui [State Key Laboratory of Silicon Materials and Department of Materials Science and Engineering, Zhejiang University, Hangzhou 310027 (China); Wu Sheng [China Academy of Engineering Physics, P.O. Box 919-71, Mianyang 621907 (China)

    2013-01-15

    Highlights: Black-Right-Pointing-Pointer The nitride layer was formed on uranium by glow plasma surface nitriding. Black-Right-Pointing-Pointer Four zones were observed in the nitride layer. Black-Right-Pointing-Pointer The chemical states of uranium, nitrogen, and oxygen were identified by AES. - Abstract: The nitride layer was formed on uranium metal by a glow plasma surface nitriding method. The structure and composition of the layer were investigated by X-ray diffraction and Auger electron spectroscopy. The nitride layer mainly consisted of {alpha}-phase U{sub 2}N{sub 3} nanocrystals with an average grain size about 10-20 nm. Four zones were identified in the layer, which were the oxide surface zone, the nitride mainstay zone, the oxide-existence interface zone, and the nitrogen-diffusion matrix zone. The gradual decrease of binding energies of uranium revealed the transition from oxide to nitride to metal states with the layer depth, while the chemical states of nitrogen and oxygen showed small variation.

  12. Simulations

    CERN Document Server

    Ngada, N M

    2015-01-01

    The complexity and cost of building and running high-power electrical systems make the use of simulations unavoidable. The simulations available today provide great understanding about how systems really operate. This paper helps the reader to gain an insight into simulation in the field of power converters for particle accelerators. Starting with the definition and basic principles of simulation, two simulation types, as well as their leading tools, are presented: analog and numerical simulations. Some practical applications of each simulation type are also considered. The final conclusion then summarizes the main important items to keep in mind before opting for a simulation tool or before performing a simulation.

  13. Identification of nitriding mechanisms in high purity reaction bonded silicon nitride

    Energy Technology Data Exchange (ETDEWEB)

    Haggerty, J.S.

    1993-03-01

    The rapid, low-temperature nitriding results from surface effects on the Si particles beginning with loss of chemisorbed H and sequential formation of thin amorphous Si nitride layers. Rapid complete conversion to Si[sub 3]N[sub 4] during the fast reaction can be inhibited when either too few or too many nuclei form on Si particels. Optimally, [approximately] 10 Si[sub 3]N[sub 4] nuclei form per Si particles under rapid, complete nitridation conditions. Nitridation during the slow reaction period appears to progress by both continued reaction of nonpreferred Si[sub 3]N[sub 4] growth interfaces and direct nitridation of the remaining Si/vapor interfaces.

  14. Identification of nitriding mechanisms in high purity reaction bonded silicon nitride

    Energy Technology Data Exchange (ETDEWEB)

    Haggerty, J.S.

    1993-03-01

    The rapid, low-temperature nitriding results from surface effects on the Si particles beginning with loss of chemisorbed H and sequential formation of thin amorphous Si nitride layers. Rapid complete conversion to Si{sub 3}N{sub 4} during the fast reaction can be inhibited when either too few or too many nuclei form on Si particels. Optimally, {approximately} 10 Si{sub 3}N{sub 4} nuclei form per Si particles under rapid, complete nitridation conditions. Nitridation during the slow reaction period appears to progress by both continued reaction of nonpreferred Si{sub 3}N{sub 4} growth interfaces and direct nitridation of the remaining Si/vapor interfaces.

  15. Colloidal Plasmonic Titanium Nitride Nanoparticles: Properties and Applications

    DEFF Research Database (Denmark)

    Guler, Urcan; Suslov, Sergey; Kildishev, Alexander V.

    2015-01-01

    Optical properties of colloidal plasmonic titanium nitride nanoparticles are examined with an eye on their photothermal and photocatalytic applications via transmission electron microscopy and optical transmittance measurements. Single crystal titanium nitride cubic nanoparticles with an average...

  16. Method of manufacture of atomically thin boron nitride

    Science.gov (United States)

    Zettl, Alexander K

    2013-08-06

    The present invention provides a method of fabricating at least one single layer hexagonal boron nitride (h-BN). In an exemplary embodiment, the method includes (1) suspending at least one multilayer boron nitride across a gap of a support structure and (2) performing a reactive ion etch upon the multilayer boron nitride to produce the single layer hexagonal boron nitride suspended across the gap of the support structure. The present invention also provides a method of fabricating single layer hexagonal boron nitride. In an exemplary embodiment, the method includes (1) providing multilayer boron nitride suspended across a gap of a support structure and (2) performing a reactive ion etch upon the multilayer boron nitride to produce the single layer hexagonal boron nitride suspended across the gap of the support structure.

  17. Temperature and Burnup Correlated FCCI in U-10Zr Metallic Fuel

    Energy Technology Data Exchange (ETDEWEB)

    William J. Carmack

    2012-05-01

    Metallic fuels are proposed for use in advanced sodium cooled fast reactors. The experience basis for metallic fuels is extensive and includes development and qualification of fuels for the Experimental Breeder Reactor I, the Experimental Breeder Reactor II, FERMI-I, and the Fast Flux Test Facility (FFTF) reactors. Metallic fuels provide a number of advantages over other fuel types in terms of fabricability, performance, recyclability, and safety. Key to the performance of all nuclear fuel systems is the resistance to “breach” and subsequent release of fission products and fuel constituents to the primary coolant system of the nuclear power plant. In metallic fuel, the experience is that significant fuel-cladding chemical (FCCI) interaction occurs and becomes prevalent at high power-high temperature operation and ultimately leads to fuel pin breach and failure. Empirical relationships for metallic fuel pin failure have been developed from a large body of in-pile and out of pile research, development, and experimentation. It has been found that significant in-pile acceleration of the FCCI rate is experienced over similar condition out-of-pile experiments. The study of FCCI in metallic fuels has led to the quantification of in-pile failure rates to establish an empirical time and temperature dependent failure limit for fuel elements. Up until now the understanding of FCCI layer formation has been limited to data generated in EBR-II experiments. This dissertation provides new FCCI data extracted from the MFF-series of metallic fuel irradiations performed in the FFTF. These fuel assemblies contain valuable information on the formation of FCCI in metallic fuels at a variety of temperature and burnup conditions and in fuel with axial fuel height three times longer than EBR-II experiments. The longer fuel column in the FFTF and the fuel pins examined have significantly different flux, power, temperature, and FCCI profiles than that found in similar tests conducted in

  18. Source Term Analysis for Reactor Coolant System with Consideration of Fuel Burnup

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yu Jong; Ahn, Joon Gi; Hwang, Hae Ryong [KEPCO EnC, Daejeon (Korea, Republic of)

    2015-10-15

    The radiation source terms in reactor coolant system (RCS) of pressurized water reactor (PWR) are basic design information for ALARA design such as radiation protection and shielding. Usually engineering companies own self-developed computer codes to estimate the source terms in RCS. DAMSAM and FIPCO are the codes developed by engineering companies. KEPCO E and C has developed computer code, RadSTAR, for use in the Radiation Source Term Analysis for Reactor coolant system during normal operation. The characteristics of RadSTAR are as follows. (1) RadSTAR uses fuel inventory data calculated by ORIGEN, such as ORIGEN2 or ORIGEN-S to consider effects of the fuel burnup. (2) RadSTAR estimates fission products by using finite differential method and analytic method to minimize numerical error. (3) RadSTAR enhances flexibility by adding the function to build the nuclide data library (production pathway library) for user-defined nuclides from ORIGEN data library. (4) RadSTAR consists of two modules. RadSTAR-BL is to build the nuclide data library. RadSTAR-ST is to perform numerical analysis on source terms. This paper includes descriptions on the numerical model, the buildup of nuclide data library, and the sensitivity analysis and verification of RadSTAR. KEPCO E and C developed RadSTAR to calculate source terms in RCS during normal operation. Sensitivity analysis and accuracy verification showed that RadSTAR keeps stability at Δt of 0.1 day and gives more accurate results in comparison with DAMSAM. After development, RadSTAR will replace DAMSAM. The areas, necessary to further development of RadSTAR, are addition of source term calculations for activation products and for shutdown operation.

  19. Diffusion kinetics of nitrogen in tantalum during plasma-nitriding

    Institute of Scientific and Technical Information of China (English)

    张德元; 林勤; 曾卫军; 李放; 许兰萍; 付青峰

    2001-01-01

    The activation energies of nitrogen in tantalum on plasma nitriding conditions were calculated according to the experimental data of hardness of plasma-nitriding of tantalum vs time and temperature. The activation energy calculated is 148.873±0.390  kJ/mol. The depth increasing of nitriding layer with time follows square root relation. The nitriding process of tantalum is controlled by diffusion of nitrogen atoms in tantalum solid solution.

  20. Design and construction of a prototype advanced on-line fuel burn-up monitoring system for the modular pebble bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Su, Bingjing; Hawari, Ayman, I.

    2004-03-30

    Modular Pebble Bed Reactor (MPBR) is a high temperature gas-cooled nuclear power reactor currently under study as a next generation reactor system. In addition to its inherently safe design, a unique feature of this reactor is its multi-pass fuel circulation in which the fuel pebbles are randomly loaded and continuously cycled through the core until they reach their prescribed End-of-Life burn-up limit. Unlike the situation with a conventional light water reactor, depending solely on computational methods to perform in-core fuel management for MPBR will be highly inaccurate. An on-line measurement system is needed to accurately assess whether a given pebble has reached its End-of-Life burn-up limit and thereby provide an on-line, automated go/no-go decision on fuel disposition on a pebble-by-pebble basis. This project investigated approaches to analyzing fuel pebbles in real time using gamma spectroscopy and possibly using passive neutron counting of spontaneous fission neutrons to provide the speed, accuracy, and burn-up range required for burnup determination of MPBR. It involved all phases necessary to develop and construct a burn-up monitor, including a review of the design requirements of the system, identification of detection methodologies, modeling and development of potential designs, and finally, the construction and testing of an operational detector system. Based upon the research work performed in this project, the following conclusions are made. In terms of using gamma spectrometry, two possible approaches were identified for burnup assay. The first approach is based on the measurement of the absolute activity of Cs-137. However, due to spectral interference and the need for absolute calibration of the spectrometer, the uncertainty in burnup determination using this approach was found to range from {approx} {+-}40% at beginning of life to {approx} {+-}10% at the discharge burnup. An alternative approach is to use a relative burnup indicator. In this

  1. Acceptor impurity activation in III-nitride light emitting diodes

    Energy Technology Data Exchange (ETDEWEB)

    Römer, Friedhard, E-mail: froemer@uni-kassel.de; Witzigmann, Bernd, E-mail: bernd.witzigmann@uni-kassel.de [Department of Electrical Engineering, University of Kassel, 34121 Kassel (Germany)

    2015-01-12

    In this work, the role of the acceptor doping and the acceptor activation and its impact on the internal quantum efficiency (IQE) of a Gallium Nitride (GaN) based multi-quantum well light emitting diode is studied by microscopic simulation. Acceptor impurities in GaN are subject to a high activation energy which depends on the presence of proximate dopant atoms and the electric field. A combined model for the dopant ionization and activation barrier reduction has been developed and implemented in a semiconductor carrier transport simulator. By model calculations, we demonstrate the impact of the acceptor activation mechanisms on the decay of the IQE at high current densities, which is known as the efficiency droop. A major contributor to the droop is the electron leakage which is largely affected by the acceptor doping.

  2. Localized surface phonon polariton resonances in polar gallium nitride

    Energy Technology Data Exchange (ETDEWEB)

    Feng, Kaijun, E-mail: kfeng@nd.edu; Islam, S. M.; Verma, Jai; Hoffman, Anthony J. [Department of Electrical Engineering, University of Notre Dame, Notre Dame, Indiana 46556 (United States); Streyer, William; Wasserman, Daniel [Department of Electrical and Computer Engineering, University of Illinois Urbana-Champaign, Urbana, Illinois 61801 (United States); Jena, Debdeep [Department of Electrical Engineering, University of Notre Dame, Notre Dame, Indiana 46556 (United States); School of Electrical and Computer Engineering, Cornell University, Ithaca, New York 14850 (United States)

    2015-08-24

    We demonstrate the excitation of localized surface phonon polaritons in an array of sub-diffraction pucks fabricated in an epitaxial layer of gallium nitride (GaN) on a silicon carbide (SiC) substrate. The array is characterized via polarization- and angle-dependent reflection spectroscopy in the mid-infrared, and coupling to several localized modes is observed in the GaN Reststrahlen band (13.4–18.0 μm). The same structure is simulated using finite element methods and the charge density of the modes are studied; transverse dipole modes are identified for the transverse electric and magnetic polarizations and a quadrupole mode is identified for the transverse magnetic polarization. The measured mid-infrared spectrum agrees well with numerically simulated spectra. This work could enable optoelectronic structures and devices that support surface modes at mid- and far-infrared wavelengths.

  3. Silicon nitride ceramic having high fatigue life and high toughness

    Science.gov (United States)

    Yeckley, Russell L.

    1996-01-01

    A sintered silicon nitride ceramic comprising between about 0.6 mol % and about 3.2 mol % rare earth as rare earth oxide, and between about 85 w/o and about 95 w/o beta silicon nitride grains, wherein at least about 20% of the beta silicon nitride grains have a thickness of greater than about 1 micron.

  4. [The effect of plasma nitriding on tungsten burs].

    Science.gov (United States)

    Cicciu, D; Russo, S; Grasso, C

    1989-01-01

    The authors have experimented the nitriding's effects on some cilindrical burs carbide utilized in dentistry after disamination on the applications methodics on plasma nitriding in neurosurgery, orthopedic surgery and in odontotherapy. This reacherys point out that nitriding plasma a durings increase and cutis greater capacity establish.

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

    Science.gov (United States)

    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).

  6. Innovative boron nitride-doped propellants

    Directory of Open Access Journals (Sweden)

    Thelma Manning

    2016-04-01

    Full Text Available The U.S. military has a need for more powerful propellants with balanced/stoichiometric amounts of fuel and oxidants. However, balanced and more powerful propellants lead to accelerated gun barrel erosion and markedly shortened useful barrel life. Boron nitride (BN is an interesting potential additive for propellants that could reduce gun wear effects in advanced propellants (US patent pending 2015-026P. Hexagonal boron nitride is a good lubricant that can provide wear resistance and lower flame temperatures for gun barrels. Further, boron can dope steel, which drastically improves its strength and wear resistance, and can block the formation of softer carbides. A scalable synthesis method for producing boron nitride nano-particles that can be readily dispersed into propellants has been developed. Even dispersion of the nano-particles in a double-base propellant has been demonstrated using a solvent-based processing approach. Stability of a composite propellant with the BN additive was verified. In this paper, results from propellant testing of boron nitride nano-composite propellants are presented, including closed bomb and wear and erosion testing. Detailed characterization of the erosion tester substrates before and after firing was obtained by electron microscopy, inductively coupled plasma and x-ray photoelectron spectroscopy. This promising boron nitride additive shows the ability to improve gun wear and erosion resistance without any destabilizing effects to the propellant. Potential applications could include less erosive propellants in propellant ammunition for large, medium and small diameter fire arms.

  7. Nitridation of chromium powder in ammonia atmosphere

    Institute of Scientific and Technical Information of China (English)

    Ling Li; Qiang Zhen; Rong Li

    2015-01-01

    CrN powder was synthesized by nitriding Cr metal in ammonia gas flow, and its chemical reaction mechanism and nitridation process were studied. Through thermodynamic calculations, the Cr−N−O predominance diagrams were constructed for different tempera-tures. Chromium nitride formed at 700−1200°C under relatively higher nitrogen and lower oxygen partial pressures. Phases in the products were then investigated using X-ray diffraction (XRD), and the Cr2N content varied with reaction temperature and holding time. The results indicate that the Cr metal powder nitridation process can be explained by a diffusion model. Further, Cr2N formed as an intermediate product because of an incomplete reaction, which was observed by high-resolution transmission electron microscopy (HRTEM). After nitriding at 1000°C for 20 h, CrN powder with an average grain size of 63 nm was obtained, and the obtained sample was analyzed by using a scanning electron microscope (SEM).

  8. Innovative boron nitride-doped propellants

    Institute of Scientific and Technical Information of China (English)

    Thelma MANNING; Henry GRAU; Paul MATTER; Michael BEACHY; Christopher HOLT; Samuel SOPOK; Richard FIELD; Kenneth KLINGAMAN; Michael FAIR; John BOLOGNINI; Robin CROWNOVER; Carlton P. ADAM; Viral PANCHAL; Eugene ROZUMOV

    2016-01-01

    The U.S. military has a need for more powerful propellants with balanced/stoichiometric amounts of fuel and oxidants. However, balanced and more powerful propellants lead to accelerated gun barrel erosion and markedly shortened useful barrel life. Boron nitride (BN) is an interesting potential additive for propellants that could reduce gun wear effects in advanced propellants (US patent pending 2015-026P). Hexagonal boron nitride is a good lubricant that can provide wear resistance and lower flame temperatures for gun barrels. Further, boron can dope steel, which drastically improves its strength and wear resistance, and can block the formation of softer carbides. A scalable synthesis method for producing boron nitride nano-particles that can be readily dispersed into propellants has been developed. Even dispersion of the nano-particles in a double-base propellant has been demonstrated using a solvent-based processing approach. Stability of a composite propellant with the BN additive was verified. In this paper, results from propellant testing of boron nitride nano-composite propellants are presented, including closed bomb and wear and erosion testing. Detailed characterization of the erosion tester substrates before and after firing was obtained by electron microscopy, inductively coupled plasma and x-ray photoelectron spectroscopy. This promising boron nitride additive shows the ability to improve gun wear and erosion resistance without any destabilizing effects to the propellant. Potential applications could include less erosive propellants in propellant ammunition for large, medium and small diameter fire arms.

  9. Development of Monteburns: A Code That Links MCNP and ORIGEN2 in an Automated Fashion for Burnup Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Holly R. Trellue

    1998-12-01

    Monteburns is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code 0RIGEN2. Monteburns produces many criticality and burnup computational parameters based on material feed/removal specifications, power(s), and time intervals. This code processes input from the user indicating the system geometry, initial material compositions, feed/removal, and other code-specific parameters. Results from MCNP, 0RIGEN2, and other calculations are then output successively as the code runs. The principle function of monteburns is to first transfer one-group cross sections and fluxes from MCNP to 0RIGEN2, and then transfer the resulting material compositions (after irradiation and/or decay) from 0RIGEN2 back to MCNP in a repeated, cyclic fashion. The main requirement of the code is that the user have a working MCNP input file and other input parameters; all interaction with 0RIGEN2 and other calculations are performed by monteburns. This report presents the results obtained from the benchmarking of monteburns to measured and previously obtained data from traditional Light Water Reactor systems. The majority of the differences seen between the two were less than five percent. These were primarily a result of variances in cross sections between MCNP, cross section libraries used by other codes, and observed values. With this understanding, this code can now be used with confidence for burnup calculations in three-dimensional systems. It was designed for use in the Accelerator Transmutation of Waste project at Los Alamos National Laboratory but is also being applied to the analysis of isotopic production/destruction of transuranic actinides in a reactor system. The code has now been shown to sufficiently support these calculations.

  10. Single gallium nitride nanowire lasers.

    Science.gov (United States)

    Johnson, Justin C; Choi, Heon-Jin; Knutsen, Kelly P; Schaller, Richard D; Yang, Peidong; Saykally, Richard J

    2002-10-01

    There is much current interest in the optical properties of semiconductor nanowires, because the cylindrical geometry and strong two-dimensional confinement of electrons, holes and photons make them particularly attractive as potential building blocks for nanoscale electronics and optoelectronic devices, including lasersand nonlinear optical frequency converters. Gallium nitride (GaN) is a wide-bandgap semiconductor of much practical interest, because it is widely used in electrically pumped ultraviolet-blue light-emitting diodes, lasers and photodetectors. Recent progress in microfabrication techniques has allowed stimulated emission to be observed from a variety of GaN microstructures and films. Here we report the observation of ultraviolet-blue laser action in single monocrystalline GaN nanowires, using both near-field and far-field optical microscopy to characterize the waveguide mode structure and spectral properties of the radiation at room temperature. The optical microscope images reveal radiation patterns that correlate with axial Fabry-Perot modes (Q approximately 10(3)) observed in the laser spectrum, which result from the cylindrical cavity geometry of the monocrystalline nanowires. A redshift that is strongly dependent on pump power (45 meV microJ x cm(-2)) supports the idea that the electron-hole plasma mechanism is primarily responsible for the gain at room temperature. This study is a considerable advance towards the realization of electron-injected, nanowire-based ultraviolet-blue coherent light sources.

  11. Nonlinear conductivity in silicon nitride

    Science.gov (United States)

    Tuncer, Enis

    2017-08-01

    To better comprehend electrical silicon-package interaction in high voltage applications requires full characterization of the electrical properties of dielectric materials employed in wafer and package level design. Not only the packaging but wafer level dielectrics, i.e. passivation layers, would experience high electric fields generated by the voltage applied pads. In addition the interface between the passivation layer and a mold compound might develop space charge because of the mismatch in electrical properties of the materials. In this contribution electrical properties of a thin silicon nitride (Si3N4) dielectric is reported as a function of temperature and electric field. The measured values later analyzed using different temperature dependent exponential expressions and found that the Mott variable range hopping conduction model was successful to express the data. A full temperature/electric field dependency of conductivity is generated. It was found that the conduction in Si3N4 could be expressed like a field ionization or Fowler-Nordheim mechanism.

  12. Study of the triton-burnup process in different JET scenarios using neutron monitor based on CVD diamond

    Science.gov (United States)

    Nemtsev, G.; Amosov, V.; Meshchaninov, S.; Popovichev, S.; Rodionov, R.

    2016-11-01

    We present the results of analysis of triton burn-up process using the data from diamond detector. Neutron monitor based on CVD diamond was installed in JET torus hall close to the plasma center. We measure the part of 14 MeV neutrons in scenarios where plasma current varies in a range of 1-3 MA. In this experiment diamond neutron monitor was also able to detect strong gamma bursts produced by runaway electrons arising during the disruptions. We can conclude that CVD diamond detector will contribute to the study of fast particles confinement and help predict the disruption events in future tokamaks.

  13. Accident source terms for pressurized water reactors with high-burnup cores calculated using MELCOR 1.8.5.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Powers, Dana Auburn; Ashbaugh, Scott G.; Leonard, Mark Thomas; Longmire, Pamela

    2010-04-01

    In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in this study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs2MoO4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU

  14. Formation and control of stoichiometric hafnium nitride thin films by direct sputtering of hafnium nitride target

    CERN Document Server

    Gotoh, Y; Ishikawa, J; Liao, M Y

    2003-01-01

    Hafnium nitride thin films were prepared by radio-frequency sputter deposition with a hafnium nitride target. Deposition was performed with various rf powers, argon pressures, and substrate temperatures, in order to investigate the influences of these parameters on the film properties, particularly the nitrogen composition. It was found that stoichiometric hafnium nitride films were formed at an argon gas pressure of less than 2 Pa, irrespective of the other deposition parameters within the range investigated. Maintaining the nitrogen composition almost stoichiometric, orientation, stress, and electrical resistivity of the films could be controlled with deposition parameters. (author)

  15. Fracture resistance of surface-nitrided zirconia

    Energy Technology Data Exchange (ETDEWEB)

    Feder, A.; Casellas, D.; Llanes, L.; Anglada, M. [Universidad Politecnica de Cataluna, Barcelona (Spain). Dept. of Material Science and Metallurgy

    2002-07-01

    Heat treatments have been conducted at 1650 C for 2 hours in Y-TZP stabilised with 2.5% molar of yttria in two different environments: in air and in nitrogen gas with the specimens embedded in a zirconium nitride powder bed. Relevant microstructural changes were induced by these heat treatments. It is highlighted the formation of a nitrided surface layer of about 400 {mu}m in thickness. Such layer has clear microstructural differences with respect to the bulk, and is formed by different sublayers with cubic and tetragonal phases with distinct degrees of transformability, as revealed by XRD and Raman spectroscopy. The fracture toughness and the hardness of the nitrided surface layer are higher than for the original Y-TZP. (orig.)

  16. Study of aluminum nitride precipitation in Fe- 3%Si steel

    Directory of Open Access Journals (Sweden)

    F.L. Alcântara

    2013-01-01

    Full Text Available For good performance of electrical steels it is necessary a high magnetic induction and a low power loss when submitted to cyclic magnetization. A fine dispersion of precipitates is a key requirement in the manufacturing process of Fe- 3%Si grain oriented electrical steel. In the production of high permeability grain oriented steel precipitate particles of copper and manganese sulphides and aluminium nitride delay normal grain growth during primary recrystallization, causing preferential growth of grains with Goss orientation during secondary recrystallization. The sulphides precipitate during the hot rolling process. The aluminium nitride particles are formed during hot rolling and the hot band annealing process. In this work AlN precipitation during hot deformation of a high permeability grain oriented 3%Si steel is examined. In the study, transfer bar samples were submitted to controlled heating, compression and cooling treatments in order to simulate a reversible hot rolling finishing. The samples were analyzed using the transmission electron microscope (TEM in order to identify the precipitates and characterize size distribution. Precipitate extraction by dissolution method and analyses by inductively coupled plasma optical emission spectrometry (ICP-OES were used to quantify the precipitation. The results allowed to describe the precipitation kinetics by a precipitation-time-temperature (PTT diagram for AlN formation during hot rolling.

  17. Instant release of fission products in leaching experiments with high burn-up nuclear fuels in the framework of the Euratom project FIRST- Nuclides

    Science.gov (United States)

    Lemmens, K.; González-Robles, E.; Kienzler, B.; Curti, E.; Serrano-Purroy, D.; Sureda, R.; Martínez-Torrents, A.; Roth, O.; Slonszki, E.; Mennecart, T.; Günther-Leopold, I.; Hózer, Z.

    2017-02-01

    The instant release of fission products from high burn-up UO2 fuels and one MOX fuel was investigated by means of leach tests. The samples covered PWR and BWR fuels at average rod burn-up in the range of 45-63 GWd/tHM and included clad fuel segments, fuel segments with opened cladding, fuel fragments and fuel powder. The tests were performed with sodium chloride - bicarbonate solutions under oxidizing conditions and, for one test, in reducing Ar/H2 atmosphere. The iodine and cesium release could be partially explained by the differences in sample preparation, leading to different sizes and properties of the exposed surface areas. Iodine and cesium releases tend to correlate with FGR and linear power rating, but the scatter of the data is significant. Although the gap between the fuel and the cladding was closed in some high burn-up samples, fissures still provide possible preferential transport pathways.

  18. Efficient and Accurate Calculation of Burnup Problems with Short-Lived Nuclides by a Krylov Subspace Method with the Newton Divided Difference

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yoon Hee; Cho, Nam Zin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-05-15

    Nowadays lattice physics codes tend to utilize a detailed burnup chain including short-lived nuclides in order to perform more accurate burnup calculations. But, since production codes, for example, ORIGEN2, take account of nuclides which have relatively long half-life, it is inappropriate for such detailed burnup chain calculation. To enhance that drawback, many matrix exponential calculation methods have been developed. Recently, a Krylov subspace method with the PADE approximation was used. In this paper, a Krylov subspace method based on spectral decomposition property of the matrix function theory with the Newton divided difference (NDD) is introduced. It is tested with a sample problem and compared with simple Taylor expansion method

  19. Chemical states of fission products in irradiated (U 0.3Pu 0.7)C 1+ x fuel at high burn-ups

    Science.gov (United States)

    Agarwal, Renu; Venugopal, V.

    2006-12-01

    The chemical states of fission products have been theoretically determined for the irradiated carbide fuel of Fast Breeder Test Reactor (FBTR) at Kalpakkam, India, at different burn-ups. The SOLGASMIX-PV computer code was used to determine the equilibrium chemical composition of the fuel. The system was assumed to be composed of a gaseous phase at one atmosphere pressure, and various solid phases. The distribution of elements in these phases and their chemical states at different temperatures were calculated as a function of burn-up. The FBTR fuel, (U 0.3Pu 0.7)C 1+ x, was loaded with C/M values in the range, 1.03-1.06. The present calculations indicated that even for the lowest starting C/M of 1.03 in the FBTR fuel, the liquid metal phase of (U, Pu), should not appear at a burn-up as high as 150 GWd/t.

  20. Nitride Fuel Development at the INL

    Energy Technology Data Exchange (ETDEWEB)

    W.E. Windes

    2007-06-01

    A new method for fabricating nitride-based fuels for nuclear applications is under development at the Idaho National Laboratory (INL). A primary objective of this research is the development of a process that could be operated as an automated or semi-automated technique reducing costs, worker doses, and eventually improving the final product form. To achieve these goals the fabrication process utilizes a new cryo-forming technique to produce microspheres formed from sub-micron oxide powder to improve material handling issues, yield rapid kinetics for conversion to nitrides, and reduced material impurity levels within the nitride compounds. The microspheres are converted to a nitride form within a high temperature particle fluidizing bed using a carbothermic process that utilizes a hydrocarbon – hydrogen - nitrogen gas mixture. A new monitor and control system using differential pressure changes in the fluidizing gas allows for real-time monitoring and control of the spouted bed reactor during conversion. This monitor and control system can provide real-time data that is used to control the gas flow rates, temperatures, and gas composition to optimize the fluidization of the particle bed. The small size (0.5 µm) of the oxide powders in the microspheres dramatically increases the kinetics of the conversion process yielding reduced process times and temperatures. Initial studies using surrogate ZrO2 powder have yielded conversion efficiencies of 90 -95 % nitride formation with only small levels of oxide and carbide contaminants present. Further studies are being conducted to determine optimal gas mixture ratios, process time, and temperature range for providing complete conversion to a nitride form.

  1. Precipitate-Accommodated Plasma Nitriding for Aluminum Alloys

    Institute of Scientific and Technical Information of China (English)

    Patama Visittipitukul; Tatsuhiko Aizawa; Hideyuki Kuwahara

    2004-01-01

    Reliable surface treatment has been explored to improve the strength and wear resistance of aluminum alloy parts in automotives. Long duration time as well as long pre-sputtering time are required for plasma nitriding of aluminum or its alloys only with the thickness of a few micrometers. New plasma inner nitriding is proposed to realize the fast-rate nitriding of aluminum alloys. Al-6Cu alloy is employed as a targeting material in order to demonstrate the effectiveness of this plasma nitriding. Mechanism of fast-rate nitriding process is discussed with consideration of the role of Al2Cu precipitates.

  2. Low pressure growth of cubic boron nitride films

    Science.gov (United States)

    Ong, Tiong P. (Inventor); Shing, Yuh-Han (Inventor)

    1997-01-01

    A method for forming thin films of cubic boron nitride on substrates at low pressures and temperatures. A substrate is first coated with polycrystalline diamond to provide a uniform surface upon which cubic boron nitride can be deposited by chemical vapor deposition. The cubic boron nitride film is useful as a substitute for diamond coatings for a variety of applications in which diamond is not suitable. any tetragonal or hexagonal boron nitride. The cubic boron nitride produced in accordance with the preceding example is particularly well-suited for use as a coating for ultra hard tool bits and abrasives, especially those intended to use in cutting or otherwise fabricating iron.

  3. Microbial adherence to a nonprecious alloy after plasma nitriding process.

    Science.gov (United States)

    Sonugelen, Mehmet; Destan, Uhmut Iyiyapici; Lambrecht, Fatma Yurt; Oztürk, Berran; Karadeniz, Süleyman

    2006-01-01

    To investigate the microbial adherence to the surfaces of a nonprecious metal alloy after plasma nitriding. The plasma-nitriding process was performed to the surfaces of metals prepared from a nickel-chromium alloy. The microorganisms were labeled with technetium-99m. After the labeling procedure, 60 metal disks were treated with a microorganism for each use. The results revealed that the amount of adherence of all microorganisms on surfaces was changed by plasma-nitriding process; adherence decreased substantially (P plasma nitriding time were not significant (P> .05) With the plasma-nitriding process, the surface properties of nonprecious metal alloys can be changed, leading to decreased microbial adherence.

  4. Alkaline Capacitors Based on Nitride Nanoparticles

    Science.gov (United States)

    Aldissi, Matt

    2003-01-01

    High-energy-density alkaline electrochemical capacitors based on electrodes made of transition-metal nitride nanoparticles are undergoing development. Transition- metal nitrides (in particular, Fe3N and TiN) offer a desirable combination of high electrical conductivity and electrochemical stability in aqueous alkaline electrolytes like KOH. The high energy densities of these capacitors are attributable mainly to their high capacitance densities, which, in turn, are attributable mainly to the large specific surface areas of the electrode nanoparticles. Capacitors of this type could be useful as energy-storage components in such diverse equipment as digital communication systems, implanted medical devices, computers, portable consumer electronic devices, and electric vehicles.

  5. A diffusion-precipitation model for gaseous nitriding of Fe-2 wt.% V alloy

    Energy Technology Data Exchange (ETDEWEB)

    Kouba, R., E-mail: r_kouba11@yahoo.fr [Departement SDM, Laboratoire de Technologie des Materiaux, Faculte de Genie Mecanique et Genie des Procedes, USTHB, BP 32 El-Alia, 16111 Alger (Algeria); Keddam, M. [Departement SDM, Laboratoire de Technologie des Materiaux, Faculte de Genie Mecanique et Genie des Procedes, USTHB, BP 32 El-Alia, 16111 Alger (Algeria); Djeghlal, M.E. [Laboratoire LSGM, Departement de Metallurgie, Ecole Nationale Polytechnique, 10 Avenue Hassen Badi, BP 182-16200 El Harrach (Algeria)

    2012-09-25

    Highlights: Black-Right-Pointing-Pointer Simulation of binary Fe-V nitriding was realized on the diffusion zone. Black-Right-Pointing-Pointer The model takes into account nitrogen diffusion in ferrite and VN precipitation. Black-Right-Pointing-Pointer VN precipitation was considered via thermodynamic equilibrium calculation. Black-Right-Pointing-Pointer The model predicts nitrogen profile and highlights nitrogen excess phenomenon. Black-Right-Pointing-Pointer The model was validated by using experimental data available on literature. - Abstract: A diffusion-precipitation model for gaseous nitriding of a Fe-2 wt.% V binary alloy has been presented. The nitriding treatment is assumed to be completely realized in the ferritic zone. The model takes into account both nitrogen diffusion and vanadium nitride precipitation. The VN precipitation was obtained with the assumption that it exists a local thermodynamic equilibrium between the matrix phase and the precipitate. The thermodynamic equilibrium calculations are based on Gibbs energies minimization, which were performed by the Thermocalc software. The suggested model allowed the prediction of the nitrogen profiles, and also takes into account the nitrogen excess phenomenon. This nitrogen excess has been explained by the presence of iron atoms within the precipitate. The theoretical results of the model have been compared to the experimental data given in the literature. A good agreement was then noticed between the experimental data and the numerical results.

  6. Boron Nitride Nanotubes for Spintronics

    Directory of Open Access Journals (Sweden)

    Kamal B. Dhungana

    2014-09-01

    Full Text Available With the end of Moore’s law in sight, researchers are in search of an alternative approach to manipulate information. Spintronics or spin-based electronics, which uses the spin state of electrons to store, process and communicate information, offers exciting opportunities to sustain the current growth in the information industry. For example, the discovery of the giant magneto resistance (GMR effect, which provides the foundation behind modern high density data storage devices, is an important success story of spintronics; GMR-based sensors have wide applications, ranging from automotive industry to biology. In recent years, with the tremendous progress in nanotechnology, spintronics has crossed the boundary of conventional, all metallic, solid state multi-layered structures to reach a new frontier, where nanostructures provide a pathway for the spin-carriers. Different materials such as organic and inorganic nanostructures are explored for possible applications in spintronics. In this short review, we focus on the boron nitride nanotube (BNNT, which has recently been explored for possible applications in spintronics. Unlike many organic materials, BNNTs offer higher thermal stability and higher resistance to oxidation. It has been reported that the metal-free fluorinated BNNT exhibits long range ferromagnetic spin ordering, which is stable at a temperature much higher than room temperature. Due to their large band gap, BNNTs are also explored as a tunnel magneto resistance device. In addition, the F-BNNT has recently been predicted as an ideal spin-filter. The purpose of this review is to highlight these recent progresses so that a concerted effort by both experimentalists and theorists can be carried out in the future to realize the true potential of BNNT-based spintronics.

  7. Boron nitride nanotubes for spintronics.

    Science.gov (United States)

    Dhungana, Kamal B; Pati, Ranjit

    2014-09-22

    With the end of Moore's law in sight, researchers are in search of an alternative approach to manipulate information. Spintronics or spin-based electronics, which uses the spin state of electrons to store, process and communicate information, offers exciting opportunities to sustain the current growth in the information industry. For example, the discovery of the giant magneto resistance (GMR) effect, which provides the foundation behind modern high density data storage devices, is an important success story of spintronics; GMR-based sensors have wide applications, ranging from automotive industry to biology. In recent years, with the tremendous progress in nanotechnology, spintronics has crossed the boundary of conventional, all metallic, solid state multi-layered structures to reach a new frontier, where nanostructures provide a pathway for the spin-carriers. Different materials such as organic and inorganic nanostructures are explored for possible applications in spintronics. In this short review, we focus on the boron nitride nanotube (BNNT), which has recently been explored for possible applications in spintronics. Unlike many organic materials, BNNTs offer higher thermal stability and higher resistance to oxidation. It has been reported that the metal-free fluorinated BNNT exhibits long range ferromagnetic spin ordering, which is stable at a temperature much higher than room temperature. Due to their large band gap, BNNTs are also explored as a tunnel magneto resistance device. In addition, the F-BNNT has recently been predicted as an ideal spin-filter. The purpose of this review is to highlight these recent progresses so that a concerted effort by both experimentalists and theorists can be carried out in the future to realize the true potential of BNNT-based spintronics.

  8. Ultrahard nanotwinned cubic boron nitride.

    Science.gov (United States)

    Tian, Yongjun; Xu, Bo; Yu, Dongli; Ma, Yanming; Wang, Yanbin; Jiang, Yingbing; Hu, Wentao; Tang, Chengchun; Gao, Yufei; Luo, Kun; Zhao, Zhisheng; Wang, Li-Min; Wen, Bin; He, Julong; Liu, Zhongyuan

    2013-01-17

    Cubic boron nitride (cBN) is a well known superhard material that has a wide range of industrial applications. Nanostructuring of cBN is an effective way to improve its hardness by virtue of the Hall-Petch effect--the tendency for hardness to increase with decreasing grain size. Polycrystalline cBN materials are often synthesized by using the martensitic transformation of a graphite-like BN precursor, in which high pressures and temperatures lead to puckering of the BN layers. Such approaches have led to synthetic polycrystalline cBN having grain sizes as small as ∼14 nm (refs 1, 2, 4, 5). Here we report the formation of cBN with a nanostructure dominated by fine twin domains of average thickness ∼3.8 nm. This nanotwinned cBN was synthesized from specially prepared BN precursor nanoparticles possessing onion-like nested structures with intrinsically puckered BN layers and numerous stacking faults. The resulting nanotwinned cBN bulk samples are optically transparent with a striking combination of physical properties: an extremely high Vickers hardness (exceeding 100 GPa, the optimal hardness of synthetic diamond), a high oxidization temperature (∼1,294 °C) and a large fracture toughness (>12 MPa m(1/2), well beyond the toughness of commercial cemented tungsten carbide, ∼10 MPa m(1/2)). We show that hardening of cBN is continuous with decreasing twin thickness down to the smallest sizes investigated, contrasting with the expected reverse Hall-Petch effect below a critical grain size or the twin thickness of ∼10-15 nm found in metals and alloys.

  9. Growth of gallium nitride and indium nitride nanowires on conductive and flexible carbon cloth substrates.

    Science.gov (United States)

    Yang, Yi; Ling, Yichuan; Wang, Gongming; Lu, Xihong; Tong, Yexiang; Li, Yat

    2013-03-07

    We report a general strategy for synthesis of gallium nitride (GaN) and indium nitride (InN) nanowires on conductive and flexible carbon cloth substrates. GaN and InN nanowires were prepared via a nanocluster-mediated growth method using a home built chemical vapor deposition (CVD) system with Ga and In metals as group III precursors and ammonia as a group V precursor. Electron microscopy studies reveal that the group III-nitride nanowires are single crystalline wurtzite structures. The morphology, density and growth mechanism of these nanowires are determined by the growth temperature. Importantly, a photoelectrode fabricated by contacting the GaN nanowires through a carbon cloth substrate shows pronounced photoactivity for photoelectrochemical water oxidation. The ability to synthesize group III-nitride nanowires on conductive and flexible substrates should open up new opportunities for nanoscale photonic, electronic and electrochemical devices.

  10. Experimental studies of superhard materials carbon nitride CNx prepared by ion-beam synthesis method

    Institute of Scientific and Technical Information of China (English)

    辛火平; 林成鲁; 许华平; 邹世昌; 石晓红; 吴兴龙; 朱宏; P.L.FHemment

    1996-01-01

    Formation of superhard materials carbon nitride CNt by using ion-beam synthesis method is reported.100-keV high-dose N+ ions were implanted into carbon thin films at different temperatures.The samples were evaluated by X-ray photoelectron spectroscopy (XPS),Fourier transformation-infrared absorption spectroscopy (FTIR),Raman spectroscopy,cross-sectional transmission electron microscopy (XTEM),Rutherford backscattering spectroscopy (RBS).X-ray diffraction analysis (XRD) and Vickers microhardness measurement.The results show that the buried carbon nitride CN> layer has been successfully formed by using 100-keV high-dose N+ ions implantation into carbon thin film.Implantation of reactive ions into silicon (IRIS) computer program has been used to simulate the formation of the buried β-C3N4 layer as N+ ions are implanted into carbon.A good agreement between experimental measurements and IRIS simulation is found.

  11. Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

    Science.gov (United States)

    Afifah, Maryam; Miura, Ryosuke; Su'ud, Zaki; Takaki, Naoyuki; Sekimoto, H.

    2015-09-01

    Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don't need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design. The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.

  12. Benchmark calculation of SCALE-PC 4.3 CSAS6 module and burnup credit criticality analysis

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Hee Sung; Ro, Seong Gy; Shin, Young Joon; Kim, Ik Soo [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-12-01

    Calculation biases of SCALE-PC CSAS6 module for PWR spent fuel, metallized spent fuel and solution of nuclear materials have been determined on the basis of the benchmark to be 0.01100, 0.02650 and 0.00997, respectively. With the aid of the code system, nuclear criticality safety analysis for the spent fuel storage pool has been carried out to determine the minimum burnup of spent fuel required for safe storage. The criticality safety analysis is performed using three types of isotopic composition of spent fuel: ORIGEN2-calculated isotopic compositions; the conservative inventory obtained from the multiplication of ORIGEN2-calculated isotopic compositions by isotopic correction factors; the conservative inventory of only U, Pu and {sup 241}Am. The results show that the minimum burnup for three cases are 990,6190 and 7270 MWd/tU, respectively in the case of 5.0 wt% initial enriched spent fuel. (author). 74 refs., 68 figs., 35 tabs.

  13. Accident source terms for light-water nuclear power plants using high-burnup or MOX fuel.

    Energy Technology Data Exchange (ETDEWEB)

    Salay, Michael (U.S. Nuclear Regulatory Commission, Washington, D.C.); Gauntt, Randall O.; Lee, Richard Y. (U.S. Nuclear Regulatory Commission, Washington, D.C.); Powers, Dana Auburn; Leonard, Mark Thomas

    2011-01-01

    Representative accident source terms patterned after the NUREG-1465 Source Term have been developed for high burnup fuel in BWRs and PWRs and for MOX fuel in a PWR with an ice-condenser containment. These source terms have been derived using nonparametric order statistics to develop distributions for the timing of radionuclide release during four accident phases and for release fractions of nine chemical classes of radionuclides as calculated with the MELCOR 1.8.5 accident analysis computer code. The accident phases are those defined in the NUREG-1465 Source Term - gap release, in-vessel release, ex-vessel release, and late in-vessel release. Important differences among the accident source terms derived here and the NUREG-1465 Source Term are not attributable to either fuel burnup or use of MOX fuel. Rather, differences among the source terms are due predominantly to improved understanding of the physics of core meltdown accidents. Heat losses from the degrading reactor core prolong the process of in-vessel release of radionuclides. Improved understanding of the chemistries of tellurium and cesium under reactor accidents changes the predicted behavior characteristics of these radioactive elements relative to what was assumed in the derivation of the NUREG-1465 Source Term. An additional radionuclide chemical class has been defined to account for release of cesium as cesium molybdate which enhances molybdenum release relative to other metallic fission products.

  14. Neutron Transport and Nuclear Burnup Analysis for the Laser Inertial Confinement Fusion-Fission Energy (LIFE) Engine

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, K J; Latkowski, J F; Abbott, R P; Boyd, J K; Powers, J J; Seifried, J E

    2008-10-24

    Lawrence Livermore National Laboratory is currently developing a hybrid fusion-fission nuclear energy system, called LIFE, to generate power and burn nuclear waste. We utilize inertial confinement fusion to drive a subcritical fission blanket surrounding the fusion chamber. It is composed of TRISO-based fuel cooled by the molten salt flibe. Low-yield (37.5 MJ) targets and a repetition rate of 13.3 Hz produce a 500 MW fusion source that is coupled to the subcritical blanket, which provides an additional gain of 4-8, depending on the fuel. In the present work, we describe the neutron transport and nuclear burnup analysis. We utilize standard analysis tools including, the Monte Carlo N-Particle (MCNP) transport code, ORIGEN2 and Monteburns to perform the nuclear design. These analyses focus primarily on a fuel composed of depleted uranium not requiring chemical reprocessing or enrichment. However, other fuels such as weapons grade plutonium and highly-enriched uranium are also under consideration. In addition, we have developed a methodology using {sup 6}Li as a burnable poison to replace the tritium burned in the fusion targets and to maintain constant power over the lifetime of the engine. The results from depleted uranium analyses suggest up to 99% burnup of actinides is attainable while maintaining full power at 2GW for more than five decades.

  15. Negative differential resistance in an (8, 0) carbon/boron nitride nanotube heterojunction*

    Institute of Scientific and Technical Information of China (English)

    Song Jiuxu; Yang Yintang; Liu Hongxia; Guo Lixin

    2011-01-01

    Using the method combined non-equilibrium Green's function with density functional theory, the electronic transport properties of an (8, 0) carbon/boron nitride nanotube heterojunction coupled to Au electrodes were investigated. In the current voltage characteristic of the heterojunction, negative differential resistance was found under positive and negative bias, which is the variation of the localization for corresponding molecular orbital caused by the applied bias voltage These results are meaningful to modeling and simulating on related electronic devices.

  16. Dispersion engineered high-Q silicon Nitride Ring-Resonators via Atomic Layer Deposition

    CERN Document Server

    Riemensberger, Johann; Herr, Tobias; Brasch, Victor; Holzwarth, Ronald; Kippenberg, Tobias J

    2012-01-01

    We demonstrate dispersion engineering of integrated silicon nitride based ring resonators through conformal coating with hafnium dioxide deposited on top of the structures via atomic layer deposition (ALD). Both, magnitude and bandwidth of anomalous dispersion can be significantly increased. All results are confirmed by high resolution frequency-comb-assisted-diode-laser spectroscopy and are in very good agreement with the simulated modification of the mode spectrum.

  17. Defects in III-nitride microdisk cavities

    Science.gov (United States)

    Ren, C. X.; Puchtler, T. J.; Zhu, T.; Griffiths, J. T.; Oliver, R. A.

    2017-03-01

    Nitride microcavities offer an exceptional platform for the investigation of light–matter interactions as well as the development of devices such as high efficiency light emitting diodes (LEDs) and low-threshold nanolasers. Microdisk geometries in particular are attractive for low-threshold lasing applications due to their ability to support high finesse whispering gallery modes (WGMs) and small modal volumes. In this article we review the effect of defects on the properties of nitride microdisk cavities fabricated using photoelectrochemical etching of an InGaN sacrificial superlattice (SSL). Threading dislocations originating from either the original GaN pseudosubstrate are shown to hinder the undercutting of microdisk cavities during the photoelectric chemical etching process resulting in whiskers of unetched material on the underside of microdisks. The unetched whiskers provide a pathway for light to escape, reducing microdisk Q-factor if located in the region occupied by the WGMs. Additionally, dislocations can affect the spectral stability of quantum dot emitters, thus hindering their effective integration in microdisk cavities. Though dislocations are clearly undesirable, the limiting factor on nitride microdisk Q-factor is expected to be internal absorption, indicating that the further optimisation of nitride microdisk cavities must incorporate both the elimination of dislocations and careful tailoring of the active region emission wavelength and background doping levels.

  18. Dynamic Characterization of Silicon Nitride Cantilevers

    NARCIS (Netherlands)

    Babaei Gavan, K.

    2009-01-01

    This thesis describes a series of experiments on dynamical characterization of silicon nitride cantilevers. These devices play an important role in micro-and nanoelectromechanical systems (MEMS and NEMS). They consist of a mechanical part, a sensor or actuator, and an electronic part for readout and

  19. Local residual stress measurements on nitride layers

    NARCIS (Netherlands)

    Mansilla, C.; Ocelik, V.; De Hosson, J. Th. M.

    2015-01-01

    In this work, local stresses in different nitrided maraging steel samples of high practical interest for industrial applications were studied through the so-called micro-slit milling method using a focused ion beam. The nitrogen concentration profiles were acquired by glow discharge optical emission

  20. Nitridation of silicon by nitrogen neutral beam

    Energy Technology Data Exchange (ETDEWEB)

    Hara, Yasuhiro, E-mail: yasuhirohara2002@yahoo.co.jp [Organization for Research and Development of Innovative Science and Technology, Kansai University, Yamate-cho 3-3-35, Suita 564-8680, Osaka (Japan); Shimizu, Tomohiro; Shingubara, Shoso [Department of Mechanical Engineering, Faculty of Engineering Science, Kansai University, Yamate-cho 3-3-35, Suita 564-8680, Osaka (Japan)

    2016-02-15

    Graphical abstract: - Highlights: • Nitrided silicon was formed by nitrogen neutral beam at room temperature. • Si{sub 3}N{sub 4} layer was formed at the acceleration voltage more than 20 V. • Formed Si{sub 3}N{sub 4} layer show the effective as the passivation film in the wet etching process. - Abstract: Silicon nitridation was investigated at room temperature using a nitrogen neutral beam (NB) extracted at acceleration voltages of less than 100 V. X-ray photoelectron spectroscopy (XPS) analysis confirmed the formation of a Si{sub 3}N{sub 4} layer on a Si (1 0 0) substrate when the acceleration voltage was higher than 20 V. The XPS depth profile indicated that nitrogen diffused to a depth of 36 nm for acceleration voltages of 60 V and higher. The thickness of the silicon nitrided layer increased with the acceleration voltages from 20 V to 60 V. Cross-sectional transmission electron microscopy (TEM) analysis indicated a Si{sub 3}N{sub 4} layer thickness of 3.1 nm was obtained at an acceleration voltage of 100 V. Moreover, it was proved that the nitrided silicon layer formed by the nitrogen NB at room temperature was effective as the passivation film in the wet etching process.

  1. Alkaline fuel cell with nitride membrane

    Science.gov (United States)

    Sun, Shen-Huei; Pilaski, Moritz; Wartmann, Jens; Letzkus, Florian; Funke, Benedikt; Dura, Georg; Heinzel, Angelika

    2017-06-01

    The aim of this work is to fabricate patterned nitride membranes with Si-MEMS-technology as a platform to build up new membrane-electrode-assemblies (MEA) for alkaline fuel cell applications. Two 6-inch wafer processes based on chemical vapor deposition (CVD) were developed for the fabrication of separated nitride membranes with a nitride thickness up to 1 μm. The mechanical stability of the perforated nitride membrane has been adjusted in both processes either by embedding of subsequent ion implantation step or by optimizing the deposition process parameters. A nearly 100% yield of separated membranes of each deposition process was achieved with layer thickness from 150 nm to 1 μm and micro-channel pattern width of 1μm at a pitch of 3 μm. The process for membrane coating with electrolyte materials could be verified to build up MEA. Uniform membrane coating with channel filling was achieved after the optimization of speed controlled dip-coating method and the selection of dimethylsulfoxide (DMSO) as electrolyte solvent. Finally, silver as conductive material was defined for printing a conductive layer onto the MEA by Ink-Technology. With the established IR-thermography setup, characterizations of MEAs in terms of catalytic conversion were performed successfully. The results of this work show promise for build up a platform on wafer-level for high throughput experiments.

  2. Gallium Nitride Crystals: Novel Supercapacitor Electrode Materials.

    Science.gov (United States)

    Wang, Shouzhi; Zhang, Lei; Sun, Changlong; Shao, Yongliang; Wu, Yongzhong; Lv, Jiaxin; Hao, Xiaopeng

    2016-05-01

    A type of single-crystal gallium nitride mesoporous membrane is fabricated and its supercapacitor properties are demonstrated for the first time. The supercapacitors exhibit high-rate capability, stable cycling life at high rates, and ultrahigh power density. This study may expand the range of crystals as high-performance electrode materials in the field of energy storage.

  3. Dynamic Characterization of Silicon Nitride Cantilevers

    NARCIS (Netherlands)

    Babaei Gavan, K.

    2009-01-01

    This thesis describes a series of experiments on dynamical characterization of silicon nitride cantilevers. These devices play an important role in micro-and nanoelectromechanical systems (MEMS and NEMS). They consist of a mechanical part, a sensor or actuator, and an electronic part for readout and

  4. Powdered Hexagonal Boron Nitride Reducing Nanoscale Wear

    Science.gov (United States)

    Chkhartishvili, L.; Matcharashvili, T.; Esiava, R.; Tsagareishvili, O.; Gabunia, D.; Margiev, B.; Gachechiladze, A.

    2013-05-01

    A morphology model is suggested for nano-powdered hexagonal boron nitride that can serve as an effective solid additive to liquid lubricants. It allows to estimate the specific surface, that is a hard-to-measure parameter, based on average size of powder particles. The model can be used also to control nanoscale wear processes.

  5. Physics behind Water Transport through Nanoporous Boron Nitride and Graphene.

    Science.gov (United States)

    Garnier, Ludovic; Szymczyk, Anthony; Malfreyt, Patrice; Ghoufi, Aziz

    2016-09-01

    In this work, molecular dynamics simulations were used to determine the surface tension profile of water on graphene and boron nitride (BN) multilayers and to predict water permeation through nanoporous graphene and BN membranes. For both graphene and BN multilayers, a decrease in surface tension (γ) was evidenced as the number of layers increased. This lessening in γ was shown to result from a negative surface tension contribution due to long-range wetting of water, which also contributes to lower water permeation through a two-layer membrane with respect to permeation through a monolayer. We also showed that a decrease in water surface tension on a BN monolayer with regards to graphene was at the origin of an increase in water permeation through BN. Our findings suggest that nanoporous BN membranes could be attractive candidates for desalination applications.

  6. Gallium Nitride Electrical Characteristics Extraction and Uniformity Sorting

    Directory of Open Access Journals (Sweden)

    Shyr-Long Jeng

    2015-01-01

    Full Text Available This study examined the output electrical characteristics—current-voltage (I-V output, threshold voltage, and parasitic capacitance—of novel gallium nitride (GaN power transistors. Experimental measurements revealed that both enhanced- and depletion-mode GaN field-effect transistors (FETs containing different components of identical specifications yielded varied turn-off impedance; hence, the FET quality was inconsistent. Establishing standardized electrical measurements can provide necessary information for designers, and measuring transistor electrical characteristics establishes its equivalent-circuit model for circuit simulations. Moreover, high power output requires multiple parallel power transistors, and sorting the difference between similar electrical characteristics is critical in a power system. An isolated gate driver detection method is proposed for sorting the uniformity from the option of the turn-off characteristic. In addition, an equivalent-circuit model for GaN FETs is established on the basis of the measured electrical characteristics and verified experimentally.

  7. An Electromagnetically Excited Silicon Nitride Beam Resonant Accelerometer

    Directory of Open Access Journals (Sweden)

    2009-02-01

    Full Text Available A resonant microbeam accelerometer of a novel highly symmetric structure based on MEMS bulk-silicon technology is proposed and some numerical modeling results for this scheme are presented. The accelerometer consists of two proof masses, four supporting hinges, two anchors, and a vibrating triple beam, which is clamped at both ends to the two proof masses. LPCVD silicon rich nitride is chosen as the resonant triple beam material, and parameter optimization of the triple-beam structure has been performed. The triple beam is excited and sensed electromagnetically by film electrodes located on the upper surface of the beam. Both simulation and experimental results show that the novel structure increases the scale factor of the resonant accelerometer, and ameliorates other performance issues such as cross axis sensitivity of insensitive input acceleration, etc.

  8. Spatial signal correlation from an III-nitride synaptic device

    Science.gov (United States)

    Zhang, Shuai; Zhu, Bingcheng; Shi, Zheng; Yuan, Jialei; Jiang, Yuan; Shen, Xiangfei; Cai, Wei; Yang, Yongchao; Wang, Yongjin

    2017-10-01

    The mechanism by which the external environment affects the internal nervous system is investigated via the spatial correlation of an III-nitride synaptic device, which combines in-plane and out-of-plane illumination. The InGaN/GaN multiple-quantum-well collector (MQW-collector) demonstrates a simultaneous light emission and light detection mode due to the unique property of the MQW-diode. The MQW-collector absorbs the internal incoming light and the external illumination at the same time to generate an integration of the excitatory postsynaptic voltages (EPSVs). Signal cognition can be distinctly decoded from the integrated EPSVs because the signal differences are maintained, which is in good agreement with the simulation results. These results suggest that the nervous system can simultaneously amplify the EPSV amplitude and achieve signal cognition by spatial EPSV summation, which can be further optimized to explore the connections between the internal nervous system and the external environment.

  9. Mechanical strength of boron nitride nanotube-polymer interfaces

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Xiaoming; Ke, Changhong, E-mail: xqwang@uga.edu, E-mail: cke@binghamton.edu [Department of Mechanical Engineering, State University of New York at Binghamton, Binghamton, New York 13902 (United States); Zhang, Liuyang; Wang, Xianqiao, E-mail: xqwang@uga.edu, E-mail: cke@binghamton.edu [College of Engineering, University of Georgia, Athens, Georgia 30602 (United States); Park, Cheol [NASA Langley Research Center, Hampton, Virginia 23681 (United States); Department of Mechanical and Aerospace Engineering, University of Virginia, Charlottesville, Virginia 22904 (United States); Fay, Catharine C. [NASA Langley Research Center, Hampton, Virginia 23681 (United States)

    2015-12-21

    We investigate the mechanical strength of boron nitride nanotube (BNNT) polymer interfaces by using in situ electron microscopy nanomechanical single-tube pull-out techniques. The nanomechanical measurements show that the shear strengths of BNNT-epoxy and BNNT-poly(methyl methacrylate) interfaces reach 323 and 219 MPa, respectively. Molecular dynamics simulations reveal that the superior load transfer capacity of BNNT-polymer interfaces is ascribed to both the strong van der Waals interactions and Coulomb interactions on BNNT-polymer interfaces. The findings of the extraordinary mechanical strength of BNNT-polymer interfaces suggest that BNNTs are excellent reinforcing nanofiller materials for light-weight and high-strength polymer nanocomposites.

  10. Modelling structure and properties of amorphous silicon boron nitride ceramics

    Directory of Open Access Journals (Sweden)

    Johann Christian Schön

    2011-06-01

    Full Text Available Silicon boron nitride is the parent compound of a new class of high-temperature stable amorphous ceramics constituted of silicon, boron, nitrogen, and carbon, featuring a set of properties that is without precedent, and represents a prototypical random network based on chemical bonds of predominantly covalent character. In contrast to many other amorphous materials of technological interest, a-Si3B3N7 is not produced via glass formation, i.e. by quenching from a melt, the reason being that the binary components, BN and Si3N4, melt incongruently under standard conditions. Neither has it been possible to employ sintering of μm-size powders consisting of binary nitrides BN and Si3N4. Instead, one employs the so-called sol-gel route starting from single component precursors such as TADB ((SiCl3NH(BCl2. In order to determine the atomic structure of this material, it has proven necessary to simulate the actual synthesis route.Many of the exciting properties of these ceramics are closely connected to the details of their amorphous structure. To clarify this structure, it is necessary to employ not only experimental probes on many length scales (X-ray, neutron- and electron scattering; complex NMR experiments; IR- and Raman scattering, but also theoretical approaches. These address the actual synthesis route to a-Si3B3N7, the structural properties, the elastic and vibrational properties, aging and coarsening behaviour, thermal conductivity and the metastable phase diagram both for a-Si3B3N7 and possible silicon boron nitride phases with compositions different from Si3N4: BN = 1 : 3. Here, we present a short comprehensive overview over the insights gained using molecular dynamics and Monte Carlo simulations to explore the energy landscape of a-Si3B3N7, model the actual synthesis route and compute static and transport properties of a-Si3BN7.

  11. Work Function Characterization of Potassium-Intercalated, Boron Nitride Doped Graphitic Petals

    Directory of Open Access Journals (Sweden)

    Patrick T. McCarthy

    2017-07-01

    Full Text Available This paper reports on characterization techniques for electron emission from potassium-intercalated boron nitride-modified graphitic petals (GPs. Carbon-based materials offer potentially good performance in electron emission applications owing to high thermal stability and a wide range of nanostructures that increase emission current via field enhancement. Furthermore, potassium adsorption and intercalation of carbon-based nanoscale emitters decreases work functions from approximately 4.6 eV to as low as 2.0 eV. In this study, boron nitride modifications of GPs were performed. Hexagonal boron nitride is a planar structure akin to graphene and has demonstrated useful chemical and electrical properties when embedded in graphitic layers. Photoemission induced by simulated solar excitation was employed to characterize the emitter electron energy distributions, and changes in the electron emission characteristics with respect to temperature identified annealing temperature limits. After several heating cycles, a single stable emission peak with work function of 2.8 eV was present for the intercalated GP sample up to 1,000 K. Up to 600 K, the potassium-intercalated boron nitride modified sample exhibited improved retention of potassium in the form of multiple emission peaks (1.8, 2.5, and 3.3 eV resulting in a large net electron emission relative to the unmodified graphitic sample. However, upon further heating to 1,000 K, the unmodified GP sample demonstrated better stability and higher emission current than the boron nitride modified sample. Both samples deintercalated above 1,000 K.

  12. Experimental and numerical study on plasma nitriding of AISI P20 mold steel

    Science.gov (United States)

    Nayebpashaee, N.; Vafaeenezhad, H.; Kheirandish, Sh.; Soltanieh, M.

    2016-09-01

    In this study, plasma nitriding was used to fabricate a hard protective layer on AISI P20 steel, at three process temperatures (450°C, 500°C, and 550°C) and over a range of time periods (2.5, 5, 7.5, and 10 h), and at a fixed gas N2:H2 ratio of 75vol%:25vol%. The morphology of samples was studied using optical microscopy and scanning electron microscopy, and the formed phase of each sample was determined by X-ray diffraction. The elemental depth profile was measured by energy dispersive X-ray spectroscopy, wavelength dispersive spectroscopy, and glow dispersive spectroscopy. The hardness profile of the samples was identified, and the microhardness profile from the surface to the sample center was recorded. The results show that ɛ-nitride is the dominant species after carrying out plasma nitriding in all strategies and that the plasma nitriding process improves the hardness up to more than three times. It is found that as the time and temperature of the process increase, the hardness and hardness depth of the diffusion zone considerably increase. Furthermore, artificial neural networks were used to predict the effects of operational parameters on the mechanical properties of plastic mold steel. The plasma temperature, running time of imposition, and target distance to the sample surface were all used as network inputs; Vickers hardness measurements were given as the output of the model. The model accurately reproduced the experimental outcomes under different operational conditions; therefore, it can be used in the effective simulation of the plasma nitriding process in AISI P20 steel.

  13. Experimental and theoretical evaluation of the laser-assisted machining of silicon nitride

    Science.gov (United States)

    Rozzi, Jay Christopher

    This study focused on the experimental and theoretical evaluation of the laser assisted machining (LAM) of silicon nitride ceramics. A laser assisted machining facility was constructed whose main components consist of a COsb2 laser and a CNC lathe. Surface temperature histories were first measured and compared to a transient, three-dimensional numerical simulation for a rotating silicon nitride workpiece heated by a translating laser for ranges of the workpiece rotational and laser-translation speeds, as well as the laser beam diameter and power. Excellent agreement was obtained between the experimental and predicted temperature histories. Laser assisted machining experiments on silicon nitride ceramic workpieces were completed for a wide range of operating conditions. Data for cutting forces and surface temperature histories illustrated that the lower bound for the avoidance of cutting tool and/or workpiece fracture for LAM is defined by the YSiAlON glass transition temperature (920-970sp°C). As temperatures near the cutting tool increase to values above the glass transition temperature range, the glassy phase softened, facilitating plastic deformation and, correspondingly, the production of semi-continuous or continuous chips. The silicon nitride machined workpiece surface roughness (Rsb{a}=0.39\\ mum) for LAM at the nominal operating condition was nearly equivalent to a value associated with the grinding of silicon nitride using a diamond wheel (Rsb{a}=0.2\\ mum). By examining the machined surfaces and chips, it was shown that LAM does not produce detectable sub-surface cracking or significant silicon nitride microstructure alteration, respectively. A transient, three-dimensional numerical heat transfer model of laser assisted machining was constructed, which includes a preheat phase and material removal, with the associated changes in the workplace geometry. Excellent agreement was obtained between the measured and predicted temperature histories. The strong

  14. Development of a method for xenon determination in the microstructure of high burn-up nuclear fuel[Dissertation 17527

    Energy Technology Data Exchange (ETDEWEB)

    Horvath, M. I

    2008-07-01

    In nuclear fuel, in approximately one quarter of the fissions, one of the two formed fission products is gaseous. These are mainly the noble gases xenon and krypton with isotopes of xenon contributing up to 90% of the product gases. These noble fission gases do not combine with other species, and have a low solubility in the normally used uranium oxide matrix. They can be dissolved in the fuel matrix or precipitate in nanometer-sized bubbles within the fuel grain, in micrometer-sized bubbles at the grain boundaries, and a fraction also precipitates in fuel pores, coming from fuel fabrication. A fraction of the gas can also be released into the plenum of the fuel rod. With increasing fission, and therefore burn-up, the ceramic fuel material experiences a transformation of its structure in the 'cooler' rim region of the fuel. A subdivision occurs of the original fuel grains of few microns size into thousands of small grains of sub-micron sizes. Additionally, larger pores are formed, which also leads into an increasing porosity in the fuel rim, called high burn-up structure. In this structure, only a small fraction of the fission gas remains in the matrix, the major quantity is said to accumulate in these pores. Because of this accumulation, the knowledge of the quantities of gas within these pores is of major interest in consideration to burn-up, fuel performance and especially for safety issues. In case of design based accidents, i.e. rapidly increasing temperature transients, the behavior of the fuel has to be estimated. Various analytical techniques have been used to determine the Xe concentration in nuclear fuel samples. The capabilities of EPMA (Electron Probe Micro-Analyser) and SIMS (Secondary Ion Mass Spectrometry) have been studied and provided some qualitative information, which has been used for determining Xe-matrix concentrations. First approaches combining these two techniques to estimate pore pressures have been recently reported. However

  15. Nitriding behavior of Ni and Ni-based binary alloys

    Energy Technology Data Exchange (ETDEWEB)

    Fonovic, Matej

    2015-01-15

    Gaseous nitriding is a prominent thermochemical surface treatment process which can improve various properties of metallic materials such as mechanical, tribological and/or corrosion properties. This process is predominantly performed by applying NH{sub 3}+H{sub 2} containing gas atmospheres serving as the nitrogen donating medium at temperatures between 673 K and 873 K (400 C and 600 C). NH{sub 3} decomposes at the surface of the metallic specimen and nitrogen diffuses into the surface adjacent region of the specimen whereas hydrogen remains in the gas atmosphere. One of the most important parameters characterizing a gaseous nitriding process is the so-called nitriding potential (r{sub N}) which determines the chemical potential of nitrogen provided by the gas phase. The nitriding potential is defined as r{sub N} = p{sub NH{sub 3}}/p{sub H{sub 2}{sup 3/2}} where p{sub NH{sub 3}} and p{sub H{sub 2}} are the partial pressures of the NH{sub 3} and H{sub 2} in the nitriding atmosphere. In contrast with nitriding of α-Fe where the nitriding potential is usually in the range between 0.01 and 1 atm{sup -1/2}, nitriding of Ni and Ni-based alloys requires employing nitriding potentials higher than 100 atm{sup -1/2} and even up to ∞ (nitriding in pure NH{sub 3} atmosphere). This behavior is compatible with decreased thermodynamic stability of the 3d-metal nitrides with increasing atomic number. Depending on the nitriding conditions (temperature, nitriding potential and treatment time), different phases are formed at the surface of the Ni-based alloys. By applying very high nitriding potential, formation of hexagonal Ni{sub 3}N at the surface of the specimen (known as external nitriding) leads to the development of a compound layer, which may improve tribological properties. Underneath the Ni{sub 3}N compound layer, two possibilities exist: (i) alloying element precipitation within the nitrided zone (known as internal nitriding) and/or (ii) development of metastable and

  16. Burn-up calculation of different thorium-based fuel matrixes in a thermal research reactor using MCNPX 2.6 code

    Directory of Open Access Journals (Sweden)

    Gholamzadeh Zohreh

    2014-12-01

    Full Text Available Decrease of the economically accessible uranium resources and the inherent proliferation resistance of thorium fuel motivate its application in nuclear power systems. Estimation of the nuclear reactor’s neutronic parameters during different operational situations is of key importance for the safe operation of nuclear reactors. In the present research, thorium oxide fuel burn-up calculations for a demonstrative model of a heavy water- -cooled reactor have been performed using MCNPX 2.6 code. Neutronic parameters for three different thorium fuel matrices loaded separately in the modelled thermal core have been investigated. 233U, 235U and 239Pu isotopes have been used as fissile element in the thorium oxide fuel, separately. Burn-up of three different fuels has been calculated at 1 MW constant power. 135X and 149Sm concentration variations have been studied in the modelled core during 165 days burn-up. Burn-up of thorium oxide enriched with 233U resulted in the least 149Sm and 135Xe productions and net fissile production of 233U after 165 days. The negative fuel, coolant and void reactivity of the used fuel assures safe operation of the modelled thermal core containing (233U-Th O2 matrix. Furthermore, utilisation of thorium breeder fuel demonstrates several advantages, such as good neutronic economy, 233U production and less production of long-lived α emitter high radiotoxic wastes in biological internal exposure point of view

  17. Large Core Code Evaluation Working Group Benchmark Problem Four: neutronics and burnup analysis of a large heterogeneous fast reactor. Part 1. Analysis of benchmark results. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Cowan, C.L.; Protsik, R.; Lewellen, J.W. (eds.)

    1984-01-01

    The Large Core Code Evaluation Working Group Benchmark Problem Four was specified to provide a stringent test of the current methods which are used in the nuclear design and analyses process. The benchmark specifications provided a base for performing detailed burnup calculations over the first two irradiation cycles for a large heterogeneous fast reactor. Particular emphasis was placed on the techniques for modeling the three-dimensional benchmark geometry, and sensitivity studies were carried out to determine the performance parameter sensitivities to changes in the neutronics and burnup specifications. The results of the Benchmark Four calculations indicated that a linked RZ-XY (Hex) two-dimensional representation of the benchmark model geometry can be used to predict mass balance data, power distributions, regionwise fuel exposure data and burnup reactivities with good accuracy when compared with the results of direct three-dimensional computations. Most of the small differences in the results of the benchmark analyses by the different participants were attributed to ambiguities in carrying out the regionwise flux renormalization calculations throughout the burnup step.

  18. A study of fuel failure behavior in high burnup HTGR fuel. Analysis by STRESS3 and STAPLE codes

    Energy Technology Data Exchange (ETDEWEB)

    Martin, David G.; Sawa, Kazuhiro; Ueta, Shouhei; Sumita, Junya [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2001-05-01

    In current high temperature gas-cooled reactors (HTGRs), Tri-isotropic coated fuel particles are employed as fuel. In safety design of the HTGR fuels, it is important to retain fission products within particles so that their release to primary coolant does not exceed an acceptable level. From this point of view, the basic design criteria for the fuel are to minimize the failure fraction of as-fabricated fuel coating layers and to prevent significant additional fuel failures during operation. This report attempts to model fuel behavior in irradiation tests using the U.K. codes STRESS3 and STAPLE. Test results in 91F-1A and HRB-22 capsules irradiation tests, which were carried out at the Japan Materials Testing Reactor of JAERI and at the High Flux Isotope Reactor of Oak Ridge National Laboratory, respectively, were employed in the calculation. The maximum burnup and fast neutron fluence were about 10%FIMA and 3 x 10{sup 25} m{sup -2}, respectively. The fuel for the irradiation tests was called high burnup fuel, whose target burnup and fast neutron fluence were higher than those of the first-loading fuel of the High Temperature Engineering Test Reactor. The calculation results demonstrated that if only mean fracture stress values of PyC and SiC are used in the calculation it is not possible to predict any particle failures, by which is meant when all three load bearing layers have failed. By contrast, when statistical variations in the fracture stresses and particle specifications are taken into account, as is done in the STAPLE code, failures can be predicted. In the HRB-22 irradiation test, it was concluded that the first two particles which had failed were defective in some way, but that the third and fourth failures can be accounted for by the pressure vessel model. In the 91F-1A irradiation test, the result showed that 1 or 2 particles had failed towards the end of irradiation in the upper capsule and no particles failed in the lower capsule. (author)

  19. 2D to 3D transition of polymeric carbon nitride nanosheets

    Energy Technology Data Exchange (ETDEWEB)

    Chamorro-Posada, Pedro [Dpto. de Teoría de la Señal y Comunicaciones e IT, Universidad de Valladolid, ETSI Telecomunicación, Paseo Belén 15, 47011 Valladolid (Spain); Vázquez-Cabo, José [Dpto. de Teoría de la Señal y Comunicaciones, Universidad de Vigo, ETSI Telecomunicación, Lagoas Marcosende s/n, Vigo (Spain); Sánchez-Arévalo, Francisco M. [Instituto de Investigaciones en Materiales (IIM), Universidad Nacional Autónoma de México, Apdo. Postal 70–360, Cd. Universitaria, México D.F. 04510 (Mexico); Martín-Ramos, Pablo [Dpto. de Teoría de la Señal y Comunicaciones e IT, Universidad de Valladolid, ETSI Telecomunicación, Paseo Belén 15, 47011 Valladolid (Spain); Laboratorio de Materiales Avanzados (Advanced Materials Laboratory) ETSIIAA, Universidad de Valladolid, Avenida de Madrid 44, 34004 Palencia (Spain); Martín-Gil, Jesús; Navas-Gracia, Luis M. [Laboratorio de Materiales Avanzados (Advanced Materials Laboratory) ETSIIAA, Universidad de Valladolid, Avenida de Madrid 44, 34004 Palencia (Spain); Dante, Roberto C., E-mail: rcdante@yahoo.com [Laboratorio de Materiales Avanzados (Advanced Materials Laboratory) ETSIIAA, Universidad de Valladolid, Avenida de Madrid 44, 34004 Palencia (Spain)

    2014-11-15

    The transition from a prevalent turbostratic arrangement with low planar interactions (2D) to an array of polymeric carbon nitride nanosheets with stronger interplanar interactions (3D), occurring for samples treated above 650 °C, was detected by terahertz-time domain spectroscopy (THz-TDS). The simulated 3D material made of stacks of shifted quasi planar sheets composed of zigzagged polymer ribbons, delivered a XRD simulated pattern in relatively good agreement with the experimental one. The 2D to 3D transition was also supported by the simulation of THz-TDS spectra obtained from quantum chemistry calculations, in which the same broad bands around 2 THz and 1.5 THz were found for 2D and 3D arrays, respectively. This transition was also in accordance with the tightening of the interplanar distance probably due to an interplanar π bond contribution, as evidenced also by a broad absorption around 2.6 eV in the UV–vis spectrum, which appeared in the sample treated at 650 °C, and increased in the sample treated at 700 °C. The band gap was calculated for 1D and 2D cases. The value of 3.374 eV for the 2D case is, within the model accuracy and precision, in a relative good agreement with the value of 3.055 eV obtained from the experimental results. - Graphical abstract: 2D lattice mode vibrations and structural changes correlated with the so called “2D to 3D transition”. - Highlights: • A 2D to 3D transition has been detected for polymeric carbon nitride. • THz-TDS allowed us to discover and detect the 2D to 3D transition of polymeric carbon nitride. • We propose a structure for polymeric carbon nitride confirming it with THz-TDS.

  20. Examination of Plasma Nitriding Microstructure with Addition of Rare Earths

    Institute of Scientific and Technical Information of China (English)

    张津

    2004-01-01

    Medium-carbon alloy steel was plasma nitrided with rare earths La,Ce and Nd into the nitriding chamber respectively.The nitriding layer microstructures with and without rare earths were compared using optical microscope,normal SEM and high resolution SEM,as well as TEM.It was found that the extent of the influence on plasma nitriding varies with different contents of rare earth.The effect of plasma nitriding is benefit from adding of Ce or Nd.The formation of hard and brittle phase Fe2-3N can be prevented and the butterfly-like structure can be improved by adding Ce or Nd.However,pure La may prevent the diffusion of nitrogen and the formation of iron nitride,and reduce the depth of diffusion layer.

  1. Liquid flow cells having graphene on nitride for microscopy

    Energy Technology Data Exchange (ETDEWEB)

    Adiga, Vivekananda P.; Dunn, Gabriel; Zettl, Alexander K.; Alivisatos, A. Paul

    2016-09-20

    This disclosure provides systems, methods, and apparatus related to liquid flow cells for microscopy. In one aspect, a device includes a substrate having a first and a second oxide layer disposed on surfaces of the substrate. A first and a second nitride layer are disposed on the first and second oxide layers, respectively. A cavity is defined in the first oxide layer, the first nitride layer, and the substrate, with the cavity including a third nitride layer disposed on walls of the substrate and the second oxide layer that define the cavity. A channel is defined in the second oxide layer. An inlet port and an outlet port are defined in the second nitride layer and in fluid communication with the channel. A plurality of viewports is defined in the second nitride layer. A first graphene sheet is disposed on the second nitride layer covering the plurality of viewports.

  2. The Study of Plasma Nitriding of AISI304 Stainless Steel

    Institute of Scientific and Technical Information of China (English)

    WANG Liang; JI Shi-jun; GAO Yu-zhou; SUN Jun-cai

    2004-01-01

    This paper presents results on the plasma nitriding of AISI 304 stainless steel at different temperatures in NH 3 gas. The working pressure was 100~200 Pa and the discharge voltage was 700~800V. The phase of nitrided layer formed on the surface was confirmed by X-ray diffraction. The hardness of the samples was measured by using a Vickers microhardness tester with the load of 50g. After nitriding at about 400 ℃ for two hours a nitrided layer consisting of single γN phase with thickness of 5μm was obtained. Microhardness measurements showed significant increase in the hardness from 240 HV (for untreated samples) up to 950 HV (for nitrided samples at temperature of 420℃). The phase composition, the thickness, the microstructure and the surface topography of the nitrided layer as well as its properties depend essentially on the process parameters.

  3. A Feasibility Study to Determine Cooling Time and Burnup of ATR Fuel Using a Nondestructive Technique and Three Types of Gamma-ray Detectors

    Energy Technology Data Exchange (ETDEWEB)

    Jorge Navarro; Rahmat Aryaeinejad,; David W. Nigg

    2011-05-01

    A Feasibility Study to Determine Cooling Time and Burnup of ATR Fuel Using a Nondestructive Technique1 Rahmat Aryaeinejad, Jorge Navarro, and David W Nigg Idaho National Laboratory Abstract Effective and efficient Advanced Test Reactor (ATR) fuel management require state of the art core modeling tools. These new tools will need isotopic and burnup validation data before they are put into production. To create isotopic, burn up validation libraries and to determine the setup for permanent fuel scanner system a feasibility study was perform. The study consisted in measuring short and long cooling time fuel elements at the ATR canal. Three gamma spectroscopy detectors (HPGe, LaBr3, and HPXe) and two system configurations (above and under water) were used in the feasibility study. The first stage of the study was to investigate which detector and system configuration would be better suited for different scenarios. The second stage of the feasibility study was to create burnup and cooling time calibrations using experimental isotopic data collected and ORIGEN 2.2 burnup data. The results of the study establish that a better spectra resolution is achieve with an above the water configuration and that three detectors can be used in the permanent fuel scanner system for different situations. In addition it was conclude that a number of isotopic ratios and absolute measurements could be used to predict ATR fuel burnup and cooling times. 1This work was supported by the U.S. Depart¬ment of Energy (DOE) under Battelle Energy Alliance, LLC Contract No. DE-AC07-05ID14517.

  4. Fully Coupled Modeling of Burnup-Dependent (U1- y , Pu y )O2- x Mixed Oxide Fast Reactor Fuel Performance

    Science.gov (United States)

    Liu, Rong; Zhou, Wenzhong; Zhou, Wei

    2016-03-01

    During the fast reactor nuclear fuel fission reaction, fission gases accumulate and form pores with the increase of fuel burnup, which decreases the fuel thermal conductivity, leading to overheating of the fuel element. The diffusion of plutonium and oxygen with high temperature gradient is also one of the important fuel performance concerns as it will affect the fuel material properties, power distribution, and overall performance of the fuel pin. In order to investigate these important issues, the (U1- y Pu y )O2- x fuel pellet is studied by fully coupling thermal transport, deformation, oxygen diffusion, fission gas release and swelling, and plutonium redistribution to evaluate the effects on each other with burnup-dependent models, accounting for the evolution of fuel porosity. The approach was developed using self-defined multiphysics models based on the framework of COMSOL Multiphysics to manage the nonlinearities associated with fast reactor mixed oxide fuel performance analysis. The modeling results showed a consistent fuel performance comparable with the previous results. Burnup degrades the fuel thermal conductivity, resulting in a significant fuel temperature increase. The fission gas release increased rapidly first and then steadily with the burnup increase. The fuel porosity increased dramatically at the beginning of the burnup and then kept constant as the fission gas released to the fuel free volume, causing the fuel temperature to increase. Another important finding is that the deviation from stoichiometry of oxygen affects greatly not only the fuel properties, for example, thermal conductivity, but also the fuel performance, for example, temperature distribution, porosity evolution, grain size growth, fission gas release, deformation, and plutonium redistribution. Special attention needs to be paid to the deviation from stoichiometry of oxygen in fuel fabrication. Plutonium content will also affect the fuel material properties and performance

  5. Analysis of Corrosion Residues Collected from the Aluminum Basket Rails of the High-Burnup Demonstration Cask.

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-03-01

    On September, 2015, an inspection was performed on the TN-32B cask that will be used for the high-burnup demonstration project. During the survey, wooden cribbing that had been placed within the cask eleven years earlier to prevent shifting of the basket during transport was removed, revealing two areas of residue on the aluminum basket rails, where they had contacted the cribbing. The residue appeared to be a corrosion product, and concerns were raised that similar attack could exist at more difficult-to-inspect locations in the canister. Accordingly, when the canister was reopened, samples of the residue were collected for analysis. This report presents the results of that assessment, which determined that the corrosion was due to the presence of the cribbing. The corrosion was associated with fungal material, and fungal activity likely contributed to an aggressive chemical environment. Once the cask has been cleaned, there will be no risk of further corrosion.

  6. French investigations of high burnup effect on LOCA thermomechanical behavior: Part 1. Experimental programmes in support of LOCA design methodologies

    Energy Technology Data Exchange (ETDEWEB)

    Waeckel, N. [EDF/SEPTEN Villeurbanne (France); GrandJean, C. [IPSN, Cadarache (France); Cauvin, R.; Lebuffe, C. [EDF/SCMI, Chinon (France)

    1997-01-01

    Within the framework of Burn-Up extension request, EDF, FRAMATOME, CEA and IPSN have carried out experimental programmes in order to provide the design of fuel rods under LOCA conditions with relevant data. The design methods used in France for LOCA are based on standard Appendix K methodology updated to take into account some penalties related to the actual conditions of the Nuclear Power Plant. Best-Estimate assessments are used as well. Experimental programmes concern plastic deformation and burst behavior of advanced claddings (EDGAR) and thermal shock quenching behavior of highly irradiated claddings (TAGCIR). The former reveals the important role played by the {alpha}/{beta} transformation kinetics related to advanced alloys (Niobium alloys) and the latter the significative impact of hydrogen charged during in-reactor corrosion on oxidation kinetics and failure behavior in terms of cooling rates.

  7. Verification of spectral burn-up codes on 2D fuel assemblies of the GFR demonstrator ALLEGRO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Čerba, Štefan, E-mail: stefan.cerba@stuba.sk [Slovak University of Technology in Bratislava, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava (Slovakia); Vrban, Branislav; Lüley, Jakub [Slovak University of Technology in Bratislava, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava (Slovakia); Dařílek, Petr [VUJE a.s., Okružná 5, 918 64 Trnava (Slovakia); Zajac, Radoslav, E-mail: radoslav.zajac@vuje.sk [VUJE a.s., Okružná 5, 918 64 Trnava (Slovakia); Nečas, Vladimír; Haščik, Ján [Slovak University of Technology in Bratislava, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava (Slovakia)

    2014-02-15

    Highlights: • Verification of the MCNPX, HELIOS and SCALE codes. • MOX and ceramic fuel assembly. • Gas-cooled fast reactor. • Burnup calculation. - Abstract: The gas-cooled fast reactor, which is one of the six GEN IV reactor concepts, is characterized by high operational temperatures and a hard neutron spectrum. The utilization of commonly used spectral codes, developed mainly for LWR reactors operated in the thermal/epithermal neutron spectrum, may be connected with systematic deviations since the main development effort of these codes has been focused on the thermal part of the neutron spectrum. To be able to carry out proper calculations for fast systems the used codes have to account for neutron resonances including the self-shielding effect. The presented study aims at verifying the spectral HELIOS, MCNPX and SCALE codes on the basis of depletion calculations of 2D MOX and ceramic fuel assemblies of the ALLEGRO gas-cooled fast reactor demonstrator in infinite lattice.

  8. Wide-bandgap III-Nitride based Second Harmonic Generation

    Science.gov (United States)

    2014-10-02

    Jun-2014 Approved for Public Release; Distribution Unlimited Final Report: Wide-bandgap III - Nitride based Second Harmonic Generation The views...Report: Wide-bandgap III - Nitride based Second Harmonic Generation Report Title It was demonstrated that GaN, AlGaN and AlN lateral polar structures can...research have been socialized to the III - Nitride Optoelectronics Center of Excellence (ARL SEDD) and to the 2013 ARO Staff Research Symposium and at

  9. Dynamic Multiaxial Response of a Hot-Pressed Aluminum Nitride

    Science.gov (United States)

    2012-01-05

    Dynamic Multiaxial Response of a Hot-Pressed Aluminum Nitride by Guangli Hu, C. Q. Chen, K. T. Ramesh, and J. W. McCauley ARL-RP-0487...Laboratory Aberdeen Proving Ground, MD 21005-5066 ARL-RP-0487 June 2014 Dynamic Multiaxial Response of a Hot-Pressed Aluminum Nitride...3. DATES COVERED (From - To) January 2010–January 2013 4. TITLE AND SUBTITLE Dynamic Multiaxial Response of a Hot-Pressed Aluminum Nitride 5a

  10. Pre-oxidized and nitrided stainless steel alloy foil for proton exchange membrane fuel cell bipolar plates: Part 1. Corrosion, interfacial contact resistance, and surface structure

    Science.gov (United States)

    Brady, M. P.; Wang, H.; Turner, J. A.; Meyer, H. M.; More, K. L.; Tortorelli, P. F.; McCarthy, B. D.

    Thermal (gas) nitridation of stainless steel alloys can yield low interfacial contact resistance (ICR), electrically conductive and corrosion-resistant nitride containing surface layers (Cr 2N, CrN, TiN, V 2N, VN, etc.) of interest for fuel cells, batteries, and sensors. This paper presents results of scale-up studies to determine the feasibility of extending the nitridation approach to thin 0.1 mm stainless steel alloy foils for proton exchange membrane fuel cell (PEMFC) bipolar plates. Developmental Fe-20Cr-4V alloy and type 2205 stainless steel foils were treated by pre-oxidation and nitridation to form low-ICR, corrosion-resistant surfaces. As-treated Fe-20Cr-4V foil exhibited target (low) ICR values, whereas 2205 foil suffered from run-to-run variation in ICR values, ranging up to 2× the target value. Pre-oxidized and nitrided surface structure examination revealed surface-through-layer-thickness V-nitride particles for the treated Fe-20Cr-4V, but near continuous chromia for treated 2205 stainless steel, which was linked to the variation in ICR values. Promising corrosion resistance was observed under simulated aggressive PEMFC anode- and cathode-side bipolar plate conditions for both materials, although ICR values were observed to increase. The implications of these findings for stamped bipolar plate foils are discussed.

  11. Thermally Nitrided Stainless Steels for Polymer Electrolyte Membrane Fuel Cell Bipolar Plates: Part 1 Model Ni-50Cr and Austenitic 349TM alloys

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Heli [National Renewable Energy Laboratory (NREL); Brady, Michael P [ORNL; Turner, John [National Renewable Energy Laboratory (NREL)

    2004-01-01

    Thermal nitridation of a model Ni-50Cr alloy at 1100 C for 2 h in pure nitrogen resulted in the formation of a continuous, protective CrN/Cr{sub 2}N surface layer with a low interfacial contact resistance. Application of similar nitridation parameters to an austenitic stainless steel, 349{sup TM}, however, resulted in a discontinuous mixture of discrete CrN, Cr{sub 2}N and (Cr,Fe){sub 2}N{sub 1-x} (x = 0--0.5) phase surface particles overlying an exposed {gamma} austenite-based matrix, rather than a continuous nitride surface layer. The interfacial contact resistance of the 349{sup TM} was reduced significantly by the nitridation treatment. However, in the simulated PEMFC environments (1 M H{sub 2}SO{sub 4} + 2 ppm F{sup -} solutions at 70 C sparged with either hydrogen or air), very high corrosion currents were observed under both anodic and cathodic conditions. This poor behavior was linked to the lack of continuity of the Cr-rich nitride surface formed on 349{sup TM} Issues regarding achieving continuous, protective Cr-nitride surface layers on stainless steel alloys are discussed.

  12. Theoretical investigations of compositional inhomogeneity around threading dislocations in III-nitride semiconductor alloys

    Science.gov (United States)

    Sakaguchi, Ryohei; Akiyama, Toru; Nakamura, Kohji; Ito, Tomonori

    2016-05-01

    The compositional inhomogeneity of group III elements around threading dislocations in III-nitride semiconductors are theoretically investigated using empirical interatomic potentials and Monte Carlo simulations. We find that the calculated atomic arrangements around threading dislocations in Al0.3Ga0.7N and In0.2Ga0.8N depend on the lattice strain around dislocation cores. Consequently, compositional inhomogeneity arises around edge dislocation cores to release the strain induced by dislocation cores. In contrast, the compositional inhomogeneity in screw dislocation is negligible owing to relatively small strain induced by dislocation cores compared with edge dislocation. These results indicate that the strain relief around dislocation cores is decisive in determining the atomic arrangements and resultant compositional inhomogeneity around threading dislocations in III-nitride semiconductor alloys.

  13. Numerical and Experimental Study on the Residual Stresses in the Nitrided Steel

    Science.gov (United States)

    Song, X.; Zhang, Zhi-Qian; Narayanaswamy, S.; Huang, Y. Z.; Zarinejad, M.

    2016-09-01

    In the present work, residual stresses distribution in the gas nitrided AISI 4140 sample has been studied using finite element (FE) simulation. The nitrogen concentration profile is obtained from the diffusion-controlled compound layer growth model, and nitrogen concentration controls the material volume change through phase transformation and lattice interstitials which results in residual stresses. Such model is validated through residual stress measurement technique—micro-ring-core method, which is applied to the nitriding process to obtain the residual stresses profiles in both the compound and diffusion layer. The numerical and experimental results are in good agreement with each other; they both indicate significant stress variation in the compound layer, which was not captured in previous research works due to the resolution limit of the traditional methods.

  14. First-principles study of transition-metal nitrides as diffusion barriers against Al

    Energy Technology Data Exchange (ETDEWEB)

    Mei, Zhi-Gang, E-mail: zmei@anl.gov; Yacout, Abdellatif M.; Kim, Yeon Soo; Hofman, Gerard; Stan, Marius

    2016-04-01

    Using density-functional theory based first-principles calculations we provided a comparative study of the diffusion barrier properties of TiN, ZrN, and HfN against Al for U–Mo dispersion fuel applications. We firstly examined the thermodynamic stability of these transition-metal nitrides with Al. The calculated heats of reaction show that both TiN and ZrN are thermodynamically unstable diffusion barrier materials, which might be decomposed by Al at relatively high temperatures. As a comparison, HfN is a stable diffusion barrier material for Al. To evaluate the kinetic stability of these nitride systems against Al diffusion, we investigated the diffusion mechanisms of Al in TiN, ZrN and HfN using atomic scale simulations. The effect of non-stoichiometry on the defect formation and Al migration was systematically studied.

  15. Marine corrosion protective coatings of hexagonal boron nitride thin films on stainless steel.

    Science.gov (United States)

    Husain, Esam; Narayanan, Tharangattu N; Taha-Tijerina, Jose Jaime; Vinod, Soumya; Vajtai, Robert; Ajayan, Pulickel M

    2013-05-22

    Recently, two-dimensional, layered materials such as graphene and hexagonal boron nitride (h-BN) have been identified as interesting materials for a range of applications. Here, we demonstrate the corrosion prevention applications of h-BN in marine coatings. The performance of h-BN/polymer hybrid coatings, applied on stainless steel, were evaluated using electrochemical techniques in simulated seawater media [marine media]. h-BN/polymer coating shows an efficient corrosion protection with a low corrosion current density of 5.14 × 10(-8) A/cm(2) and corrosion rate of 1.19 × 10(-3) mm/year and it is attributed to the hydrofobic, inert and dielectric nature of boron nitride. The results indicated that the stainless steel with coatings exhibited improved corrosion resistance. Electrochemical impedance spectroscopy and potentiodynamic analysis were used to propose a mechanism for the increased corrosion resistance of h-BN coatings.

  16. Review of actinide nitride properties with focus on safety aspects

    Energy Technology Data Exchange (ETDEWEB)

    Albiol, Thierry [CEA Cadarache, St Paul Lez Durance Cedex (France); Arai, Yasuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-12-01

    This report provides a review of the potential advantages of using actinide nitrides as fuels and/or targets for nuclear waste transmutation. Then a summary of available properties of actinide nitrides is given. Results from irradiation experiments are reviewed and safety relevant aspects of nitride fuels are discussed, including design basis accidents (transients) and severe (core disruptive) accidents. Anyway, as rather few safety studies are currently available and as many basic physical data are still missing for some actinide nitrides, complementary studies are proposed. (author)

  17. Magnetism induced by electrochemical nitriding on an austenitic stainless steel

    National Research Council Canada - National Science Library

    Watanabe, Takashi; Sagara, Akio; Hishinuma, Yoshimitsu; Takayama, Sadatsugu; Tanaka, Teruya; Sano, Saburo

    2015-01-01

    .... The Nitrogen diffusion layers were predominately formed at nitrogen concentration of 23 at%. The nitriding process drastically also changed its magnetic property from non-magnetic to ferromagnetic...

  18. Synthesis of Uranium nitride powders using metal uranium powders

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Jae Ho; Kim, Dong Joo; Oh, Jang Soo; Rhee, Young Woo; Kim, Jong Hun; Kim, Keon Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    Uranium nitride (UN) is a potential fuel material for advanced nuclear reactors because of their high fuel density, high thermal conductivity, high melting temperature, and considerable breeding capability in LWRs. Uranium nitride powders can be fabricated by a carbothermic reduction of the oxide powders, or the nitriding of metal uranium. The carbothermic reduction has an advantage in the production of fine powders. However it has many drawbacks such as an inevitable engagement of impurities, process burden, and difficulties in reusing of expensive N{sup 15} gas. Manufacturing concerns issued in the carbothermic reduction process can be solved by changing the starting materials from oxide powder to metals. However, in nitriding process of metal, it is difficult to obtain fine nitride powders because metal uranium is usually fabricated in the form of bulk ingots. In this study, a simple reaction method was tested to fabricate uranium nitride powders directly from uranium metal powders. We fabricated uranium metal spherical powder and flake using a centrifugal atomization method. The nitride powders were obtained by thermal treating those metal particles under nitrogen containing gas. We investigated the phase and morphology evolutions of powders during the nitriding process. A phase analysis of nitride powders was also a part of the present work.

  19. EXAFS investigation of low temperature nitrided stainless steel

    DEFF Research Database (Denmark)

    Oddershede, Jette; Christiansen, Thomas; Ståhl, Kenny

    2008-01-01

    Low temperature nitrided stainless steel AISI 316 flakes were investigated with EXAFS and X-ray diffraction analysis. The stainless steel flakes were transformed into a mixture of nitrogen expanded austenite and nitride phases. Two treatments were carried out yielding different overall nitrogen...... contents: (1) nitriding in pure NH3 and (2)nitriding in pure NH3 followed by reduction in H2. The majority of the Cr atoms in the stainless steel after treatment 1 and 2 was associated with a nitrogen–chromium bond distance comparable to that of the chemical compound CrN. The possibility of the occurrence...

  20. Comparative infrared study of silicon and germanium nitrides

    Science.gov (United States)

    Baraton, M. I.; Marchand, R.; Quintard, P.

    1986-03-01

    Silicon and germanium nitride (Si 3N 4 and Ge 3N 4) are isomorphic compounds. They have been studied in the β-phase which crystallises in the hexagonal system. The space group is P6 3/m (C 6h2). The IR transmission spectra of these two nitrides are very similar but the absorption frequencies of germanium nitride are shifted to the lower values in comparison with silicon nitride. We noted that the atomic mass effect is the only cause of this shift for the streching modes but not for the bending modes.

  1. Thermodynamics, kinetics and process control of nitriding

    DEFF Research Database (Denmark)

    Mittemeijer, Eric J.; Somers, Marcel A. J.

    1997-01-01

    As a prerequisite for the predictability of properties obtained by a nitriding treatment of iron based workpieces, the relation between the process parameters and the composition and structure of the surface layer produced must be known. At present, even the description of thermodynamic equilibrium...... of pure Fe-N phases has not been fully achieved. It is shown that taking into account the ordering of nitrogen in the epsilon and gamma' iron nitride phases leads to an improved understanding of the Fe-N phase diagram. Although consideration of thermodynamics indicates the state the system strives for...... of the International Federation for Heat Treatment and Surface Engineering held in Brighton, UK on 1-5 September 1996. (C) 1997 The Institute of Materials....

  2. Aluminum Reduction and Nitridation of Bauxite

    Institute of Scientific and Technical Information of China (English)

    ZHANG Zhikuan; ZHANG Dianwei; XU Enxia; HOU Xinmei; DONG Yanling

    2007-01-01

    The application of bauxite with low Al2O3 content has been studied in this paper and β-SiAlON has been obtained from two kinds of bauxites (Al203 content 68.08 mass% and 46.30 mass% respectively) by aluminum reduction and nitridation method.The sequence of reactions has been studied using thermal analysis (TG-DTA),X-ray diffraction (XRD) analysis and scanning electron microscopy (SEM) with EDS.Compared with carbon thermal reduction and nitridation of aluminosilicates employed presently,the reaction in the system of bauxite-Al-N2 occurs at lower temperature.β-SiAlON appears as one of the main products from 1573K and exists' stably in the range of the present experimental temperature.The microstructure of β-SiAlON obtained at 1773 K is short column with 5-10μm observed by SEM.

  3. Sheath Characteristic in ECR Plasma Nitriding

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The sheath plasma characteristics changing with the negative bias applied to the substrate during electron cyclotron resonance plasma nitriding are studied. The sheath characteristics obtained by a Langmuir single probe and an ion energy analyzer show that when the negative bias applied to the substrate is increasing, the most probable energy of ions in the sheath and the full width of half maximum of ions energy distribution increase, the thickness of the sheath also increases, whereas the saturation current of ion decreases. It has been found from the optical emission spectrum that there are strong lines of N2 and N2+. Based on our experiment results the mechanism of plasma nitriding is discussed.

  4. Thermodynamic ground states of platinum metal nitrides

    Energy Technology Data Exchange (ETDEWEB)

    Aberg, D; Sadigh, B; Crowhurst, J; Goncharov, A

    2007-10-09

    We have systematically studied the thermodynamic stabilities of various phases of the nitrides of the platinum metal elements using density functional theory. We show that for the nitrides of Rh, Pd, Ir and Pt two new crystal structures, in which the metal ions occupy simple tetragonal lattice sites, have lower formation enthalpies at ambient conditions than any previously proposed structures. The region of stability can extend up to 17 GPa for PtN{sub 2}. Furthermore, we show that according to calculations using the local density approximation, these new compounds are also thermodynamically stable at ambient pressure and thus may be the ground state phases for these materials. We further discuss the fact that the local density and generalized gradient approximations predict different values of the absolute formation enthalpies as well different relative stabilities between simple tetragonal and the pyrite or marcasite structures.

  5. Atomic-layer deposition of silicon nitride

    CERN Document Server

    Yokoyama, S; Ooba, K

    1999-01-01

    Atomic-layer deposition (ALD) of silicon nitride has been investigated by means of plasma ALD in which a NH sub 3 plasma is used, catalytic ALD in which NH sub 3 is dissociated by thermal catalytic reaction on a W filament, and temperature-controlled ALD in which only a thermal reaction on the substrate is employed. The NH sub 3 and the silicon source gases (SiH sub 2 Cl sub 2 or SiCl sub 4) were alternately supplied. For all these methods, the film thickness per cycle was saturated at a certain value for a wide range of deposition conditions. In the catalytic ALD, the selective deposition of silicon nitride on hydrogen-terminated Si was achieved, but, it was limited to only a thin (2SiO (evaporative).

  6. Group-III Nitride Field Emitters

    Science.gov (United States)

    Bensaoula, Abdelhak; Berishev, Igor

    2008-01-01

    Field-emission devices (cold cathodes) having low electron affinities can be fabricated through lattice-mismatched epitaxial growth of nitrides of elements from group III of the periodic table. Field emission of electrons from solid surfaces is typically utilized in vacuum microelectronic devices, including some display devices. The present field-emission devices and the method of fabricating them were developed to satisfy needs to reduce the cost of fabricating field emitters, make them compatible with established techniques for deposition of and on silicon, and enable monolithic integration of field emitters with silicon-based driving circuitry. In fabricating a device of this type, one deposits a nitride of one or more group-III elements on a substrate of (111) silicon or other suitable material. One example of a suitable deposition process is chemical vapor deposition in a reactor that contains plasma generated by use of electron cyclotron resonance. Under properly chosen growth conditions, the large mismatch between the crystal lattices of the substrate and the nitride causes strains to accumulate in the growing nitride film, such that the associated stresses cause the film to crack. The cracks lie in planes parallel to the direction of growth, so that the growing nitride film becomes divided into microscopic growing single-crystal columns. The outer ends of the fully-grown columns can serve as field-emission tips. By virtue of their chemical compositions and crystalline structures, the columns have low work functions and high electrical conductivities, both of which are desirable for field emission of electrons. From examination of transmission electron micrographs of a prototype device, the average column width was determined to be about 100 nm and the sharpness of the tips was determined to be characterized by a dimension somewhat less than 100 nm. The areal density of the columns was found to about 5 x 10(exp 9)/sq cm . about 4 to 5 orders of magnitude

  7. Simulation

    DEFF Research Database (Denmark)

    Gould, Derek A; Chalmers, Nicholas; Johnson, Sheena J

    2012-01-01

    Recognition of the many limitations of traditional apprenticeship training is driving new approaches to learning medical procedural skills. Among simulation technologies and methods available today, computer-based systems are topical and bring the benefits of automated, repeatable, and reliable p...... performance assessments. Human factors research is central to simulator model development that is relevant to real-world imaging-guided interventional tasks and to the credentialing programs in which it would be used.......Recognition of the many limitations of traditional apprenticeship training is driving new approaches to learning medical procedural skills. Among simulation technologies and methods available today, computer-based systems are topical and bring the benefits of automated, repeatable, and reliable...

  8. New nanoforms of carbon and boron nitride

    Energy Technology Data Exchange (ETDEWEB)

    Pokropivny, V V [Institute for Problems of Materials Science of National Academy of Sciences of Ukraine (Ukraine); Ivanovskii, A L [Institute of Solid State Chemistry, Urals Branch of the Russian Academy of Sciences, Ekaterinburg (Russian Federation)], e-mail: Ivanovskii@ihim.uran.ru

    2008-10-31

    Data on new carbon nanostructures including those based on fullerenes, nanotubes as well monolithic diamond-like nanoparticles, nanofibres, various nanocomposites, etc., published in the last decade are generalised. The experimental and theoretical data on their atomic and electronic structures, the nature of chemical bonds and physicochemical properties are discussed. These data are compared with the results obtained in studies of nanoforms of boron nitride, an isoelectronic analogue of carbon. Potential fields of applications of the new nanostructures are considered.

  9. New nanoforms of carbon and boron nitride

    Science.gov (United States)

    Pokropivny, V. V.; Ivanovskii, A. L.

    2008-10-01

    Data on new carbon nanostructures including those based on fullerenes, nanotubes as well monolithic diamond-like nanoparticles, nanofibres, various nanocomposites, etc., published in the last decade are generalised. The experimental and theoretical data on their atomic and electronic structures, the nature of chemical bonds and physicochemical properties are discussed. These data are compared with the results obtained in studies of nanoforms of boron nitride, an isoelectronic analogue of carbon. Potential fields of applications of the new nanostructures are considered.

  10. Silicon Nitride Balls For Cryogenic Bearings

    Science.gov (United States)

    Butner, Myles F.; Ng, Lillian W.

    1990-01-01

    Resistance to wear greater than that of 440C steel. Experiments show lives of ball bearings immersed in liquid nitrogen or liquid oxygen increased significantly when 440C steel balls (running on 440C steel races) replaced by balls of silicon nitride. Developed for use at high temperatures, where lubrication poor or nonexistent. Best wear life of any bearing tested to date and ball material spalls without fracturing. Plans for future tests call for use of liquid oxygen as working fluid.

  11. Silicon Nitride Antireflection Coatings for Photovoltaic Cells

    Science.gov (United States)

    Johnson, C.; Wydeven, T.; Donohoe, K.

    1984-01-01

    Chemical-vapor deposition adapted to yield graded index of refraction. Silicon nitride deposited in layers, refractive index of which decreases with distance away from cell/coating interface. Changing index of refraction allows adjustment of spectral transmittance for wavelengths which cell is most effective at converting light to electric current. Average conversion efficiency of solar cells increased from 8.84 percent to 12.63 percent.

  12. Oxygen radical functionalization of boron nitride nanosheets

    OpenAIRE

    MAY, PETER; Coleman, Jonathan; MCGOVERN, IGNATIUS; GOUNKO, IOURI; Satti, Amro

    2012-01-01

    PUBLISHED The covalent chemical functionalization of exfoliated hexagonal boron-nitride nanosheets (BNNSs) is achieved by the solution phase oxygen radical functionalization of boron atoms in the h-BN lattice. This involves a two-step procedure to initially covalently graft alkoxy groups to boron atoms and the subsequent hydrolytic defunctionalisation of the groups to yield hydroxyl-functionalized BNNSs (OH-BNNSs). Characterization of the functionalized-BNNSs using HR-TEM, Raman, UV-Vis, F...

  13. Defects in III-Nitride Microdisk Cavities

    OpenAIRE

    Ren, C. X.; Puchtler, T. J.; Zhu, T.; Griffiths, J. T.; R. A. Oliver

    2016-01-01

    This is the author accepted manuscript. It is currently under an indefinite embargo pending publication by the Institute of Physics. Nitride microcavities offer an exceptional platform for the investigation of light-matter interactions as well as the development of devices such as high efficiency light emitting diodes (LEDs) and low-threshold nanolasers. Microdisk geometries in particular are attractive for low-threshold lasing applications due to their ability to support high finesse whis...

  14. Amorphous Carbon-Boron Nitride Nanotube Hybrids

    Science.gov (United States)

    Kim, Jae Woo (Inventor); Siochi, Emilie J. (Inventor); Wise, Kristopher E. (Inventor); Lin, Yi (Inventor); Connell, John (Inventor)

    2016-01-01

    A method for joining or repairing boron nitride nanotubes (BNNTs). In joining BNNTs, the nanotube structure is modified with amorphous carbon deposited by controlled electron beam irradiation to form well bonded hybrid a-C/BNNT structures. In repairing BNNTs, the damaged site of the nanotube structure is modified with amorphous carbon deposited by controlled electron beam irradiation to form well bonded hybrid a-C/BNNT structures at the damage site.

  15. Formation and Structure of Boron Nitride Nanotubes

    Institute of Scientific and Technical Information of China (English)

    Jiang ZHANG; Zongquan LI; Jin XU

    2005-01-01

    Boron nitride (BN) nanotubes were simply synthesized by heating well-mixed boric acid, urea and iron nitrate powders at 1000℃. A small amount of BN nanowires was also obtained in the resultants. The morphological and structural characters of the BN nanostructures were studied using transmission electron microscopy. Other novel BN nanostructures, such as Y-junction nanotubes and bamboo-like nanotubes, were simultaneously observed. The growth mechanism of the BN nanotubes was discussed briefly.

  16. Wetting and infiltration of nitride bonded silicon nitride by liquid silicon

    Science.gov (United States)

    Schneider, V.; Reimann, C.; Friedrich, J.

    2016-04-01

    Nitride bonded silicon nitride (NBSN) is a promising crucible material for the repeated use in the directional solidification of multicrystalline (mc) silicon ingots for photovoltaic applications. Due to wetting and infiltration, however, silicon nitride in its initial state does not offer the desired reusability. In this work the sessile drop method is used to systematically study the wetting and infiltration behavior of NBSN after applying different oxidation procedures. It is found that the wetting of the NBSN crucible by liquid silicon can be prevented by the oxidation of the geometrical surface. The infiltration of liquid silicon into the porous crucible can be suppressed by oxygen enrichment within the volume of the NBSN, i.e. at the pore walls of the crucibles. The realized reusability of the NBSN is demonstrated by reusing a NBSN crucible six times for the directional solidification of undoped multicrystalline silicon ingots.

  17. Simulation

    CERN Document Server

    Ross, Sheldon

    2006-01-01

    Ross's Simulation, Fourth Edition introduces aspiring and practicing actuaries, engineers, computer scientists and others to the practical aspects of constructing computerized simulation studies to analyze and interpret real phenomena. Readers learn to apply results of these analyses to problems in a wide variety of fields to obtain effective, accurate solutions and make predictions about future outcomes. This text explains how a computer can be used to generate random numbers, and how to use these random numbers to generate the behavior of a stochastic model over time. It presents the statist

  18. Water adsorption on fullerene-like carbon nitride overcoats

    Energy Technology Data Exchange (ETDEWEB)

    Broitman, E. [Department of Chemical Engineering, Carnegie Mellon University, Pittsburgh, PA 15213 (United States)], E-mail: broitman@andrew.cmu.edu; Gueorguiev, G.K.; Furlan, A.; Son, N.T. [IFM, Linkoeping University, SE 581-83 Linkoeping (Sweden); Gellman, A.J. [Department of Chemical Engineering, Carnegie Mellon University, Pittsburgh, PA 15213 (United States); Stafstroem, S.; Hultman, L. [IFM, Linkoeping University, SE 581-83 Linkoeping (Sweden)

    2008-12-01

    Humidity influences the tribological performance of the head-disk interface in magnetic data storage devices. In this work we compare the uptake of water of amorphous carbon nitride (a-CN{sub x}) films, widely used as protective overcoats in computer disk drive systems, with fullerene-like carbon nitride (FL-CN{sub x}) and amorphous carbon (a-C) films. Films with thickness in the range 10-300 nm were deposited on quartz crystal substrates by reactive DC magnetron sputtering. A quartz crystal microbalance placed in a vacuum chamber was used to measure the water adsorption. Electron paramagnetic resonance (EPR) has been used to correlate water adsorption with film microstructure and surface defects (dangling bonds). Measurements indicate that the amount of adsorbed water is highest for the pure a-C films and that the FL-CN{sub x} films adsorbed less than a-CN{sub x}. EPR data correlate the lower water adsorption on FL-CN{sub x} films with a possible lack of dangling bonds on the film surface. To provide additional insight into the atomic structure of defects in the FL-CN{sub x}, a-CN{sub x} and a-C compounds, we performed first-principles calculations within the framework of Density Functional Theory. Emphasis was put on the energy cost for formation of vacancy defects and dangling bonds in relaxed systems. Cohesive energy comparison reveals that the energy cost formation for dangling bonds in different configurations is considerably higher in FL-CN{sub x} than for the amorphous films. These simulations thus confirm the experimental results showing that dangling bonds are much less likely in FL-CN{sub x} than in a-CN{sub x} and a-C films.

  19. III-nitride monolithic LED covering full RGB color gamut

    Science.gov (United States)

    El-Ghoroury, Hussein S.; Chuang, Chih-Li; Kisin, Mikhail V.

    2016-03-01

    We present numerical simulation of III-nitride monolithic multi-color LED covering full red-green-blue (RGB) color gamut. The RGB LED structure was grown at Ostendo Technologies Inc. and has been used in Ostendo proprietary Quantum Photonic Imager (QPI) device. Active region of our RGB LED incorporates specially designed intermediate carrier blocking layers (ICBLs) controlling transport of each type of carriers and subsequent carrier injection redistribution among the optically active quantum wells (QWs) with different emission wavelengths. ICBLs are proved to be essential elements of multi-color LED active region design requiring optimization both in material composition and doping level. Strong interdependence between ICBL parameters and active QW characteristics presents additional challenge to multi-color LED design. Combination of several effects was crucial for adequate simulation of RGB LED color control features. Standard drift-diffusion transport model has been appended with rate equations for dynamic QW-confined carrier populations which appear severely off-balanced from corresponding mobile carrier subsystems. QW overshoot and Auger-assisted QW depopulation were also included into the carrier kinetic model thus enhancing the non-equilibrium character of QW confined populations and supporting the mobile carrier transport across the MQW active region. For device simulation we use COMSOL-based program suit developed at Ostendo Technologies Inc.

  20. Fusion bonding of silicon nitride surfaces

    DEFF Research Database (Denmark)

    Reck, Kasper; Østergaard, Christian; Thomsen, Erik Vilain

    2011-01-01

    While silicon nitride surfaces are widely used in many micro electrical mechanical system devices, e.g. for chemical passivation, electrical isolation or environmental protection, studies on fusion bonding of two silicon nitride surfaces (Si3N4–Si3N4 bonding) are very few and highly application...... specific. Often fusion bonding of silicon nitride surfaces to silicon or silicon dioxide to silicon surfaces is preferred, though Si3N4–Si3N4 bonding is indeed possible and practical for many devices as will be shown in this paper. We present an overview of existing knowledge on Si3N4–Si3N4 bonding and new...... results on bonding of thin and thick Si3N4 layers. The new results include high temperature bonding without any pretreatment, along with improved bonding ability achieved by thermal oxidation and chemical pretreatment. The bonded wafers include both unprocessed and processed wafers with a total silicon...

  1. Gallium nitride based logpile photonic crystals.

    Science.gov (United States)

    Subramania, Ganapathi; Li, Qiming; Lee, Yun-Ju; Figiel, Jeffrey J; Wang, George T; Fischer, Arthur J

    2011-11-09

    We demonstrate a nine-layer logpile three-dimensional photonic crystal (3DPC) composed of single crystalline gallium nitride (GaN) nanorods, ∼100 nm in size with lattice constants of 260, 280, and 300 nm with photonic band gap in the visible region. This unique GaN structure is created through a combined approach of a layer-by-layer template fabrication technique and selective metal organic chemical vapor deposition (MOCVD). These GaN 3DPC exhibit a stacking direction band gap characterized by strong optical reflectance between 380 and 500 nm. By introducing a "line-defect" cavity in the fifth (middle) layer of the 3DPC, a localized transmission mode with a quality factor of 25-30 is also observed within the photonic band gap. The realization of a group III nitride 3DPC with uniform features and a band gap at wavelengths in the visible region is an important step toward realizing complete control of the electromagnetic environment for group III nitride based optoelectronic devices.

  2. Control of Defects in Aluminum Gallium Nitride ((Al)GaN) Films on Grown Aluminum Nitride (AlN) Substrates

    Science.gov (United States)

    2013-02-01

    like HEMTs . A nanolayer of AlGaN over GaN provides extra 2DEG charge density because of the piezoelectric effect of the AlGaN layer. The higher...Control of Defects in Aluminum Gallium Nitride ((Al) GaN ) Films on Grown Aluminum Nitride (AlN) Substrates by Iskander G. Batyrev, Chi-Chin Wu...Aluminum Gallium Nitride ((Al) GaN ) Films on Grown Aluminum Nitride (AlN) Substrates Iskander G. Batyrev and N. Scott Weingarten Weapons and

  3. Development of compound layer of iron (carbo)nitrides during nitriding of steel

    DEFF Research Database (Denmark)

    Ratajski, J.; Tacikowski, J.; Somers, Marcel A.J.

    2003-01-01

    The composition and phase constitution of a compound layer developing during gaseous nitriding was investigated at 853 K for three commercial steels (AISI 120, 4340 and 1090) and Armco iron. The compound layers were characterised by light optical microscopy, X-ray diffraction and electron probe...... microanalysis. The formation of the compound layer occurs along two distinct sequences: alpha-gamma prime-epsilon and/or alpha(theta)-epsilon2-gamma prime-epsilon1. The preferred sequence depends mainly on the chemical composition of steel and on the nitriding potential....

  4. Development of compound layer of iron (carbo)nitrides during nitriding of steel

    DEFF Research Database (Denmark)

    Ratajski, J.; Tacikowski, J.; Somers, Marcel A.J.

    2003-01-01

    The composition and phase constitution of a compound layer developing during gaseous nitriding was investigated at 853 K for three commercial steels (AISI 120, 4340 and 1090) and Armco iron. The compound layers were characterised by light optical microscopy, X-ray diffraction and electron probe...... microanalysis. The formation of the compound layer occurs along two distinct sequences: alpha-gamma prime-epsilon and/or alpha(theta)-epsilon2-gamma prime-epsilon1. The preferred sequence depends mainly on the chemical composition of steel and on the nitriding potential....

  5. Impurity Influence on Nitride LEDs

    Directory of Open Access Journals (Sweden)

    O.I. Rabinovich

    2014-07-01

    Full Text Available Light emitting diodes (LEDs are widely used nowadays. They are used in major parts of our life. But it is still necessary to improve their characteristics. In this paper the impurity and Indium atoms influence on the LEDs characteristics is investigated by computer simulation. Simulation was carried out in Sim Windows. The program was improved for this purpose by creating new files for AlGaInN heterostructure and devices including more than 25 basic parameters. It was found that characteristics depend on impurity and indium atoms changes a lot. The optimum impurity concentration for doping barriers between quantum wells was achieved. By varying impurity and Indium concentration the distribution in AlGaInN heterostructure LEDs characteristics could be improved.

  6. Development of a method for xenon determination in the microstructure of high burn-up nuclear fuel[Dissertation 17527

    Energy Technology Data Exchange (ETDEWEB)

    Horvath, M. I

    2008-07-01

    In nuclear fuel, in approximately one quarter of the fissions, one of the two formed fission products is gaseous. These are mainly the noble gases xenon and krypton with isotopes of xenon contributing up to 90% of the product gases. These noble fission gases do not combine with other species, and have a low solubility in the normally used uranium oxide matrix. They can be dissolved in the fuel matrix or precipitate in nanometer-sized bubbles within the fuel grain, in micrometer-sized bubbles at the grain boundaries, and a fraction also precipitates in fuel pores, coming from fuel fabrication. A fraction of the gas can also be released into the plenum of the fuel rod. With increasing fission, and therefore burn-up, the ceramic fuel material experiences a transformation of its structure in the 'cooler' rim region of the fuel. A subdivision occurs of the original fuel grains of few microns size into thousands of small grains of sub-micron sizes. Additionally, larger pores are formed, which also leads into an increasing porosity in the fuel rim, called high burn-up structure. In this structure, only a small fraction of the fission gas remains in the matrix, the major quantity is said to accumulate in these pores. Because of this accumulation, the knowledge of the quantities of gas within these pores is of major interest in consideration to burn-up, fuel performance and especially for safety issues. In case of design based accidents, i.e. rapidly increasing temperature transients, the behavior of the fuel has to be estimated. Various analytical techniques have been used to determine the Xe concentration in nuclear fuel samples. The capabilities of EPMA (Electron Probe Micro-Analyser) and SIMS (Secondary Ion Mass Spectrometry) have been studied and provided some qualitative information, which has been used for determining Xe-matrix concentrations. First approaches combining these two techniques to estimate pore pressures have been recently reported. However

  7. Phase diagrams and synthesis of cubic boron nitride

    CERN Document Server

    Turkevich, V Z

    2002-01-01

    On the basis of phase equilibria, the lowest temperatures, T sub m sub i sub n , above which at high pressures cubic boron nitride crystallization from melt solution is allowable in terms of thermodynamics have been found for a number of systems that include boron nitride.

  8. Vertical III-nitride thin-film power diode

    Energy Technology Data Exchange (ETDEWEB)

    Wierer, Jr., Jonathan; Fischer, Arthur J.; Allerman, Andrew A.

    2017-03-14

    A vertical III-nitride thin-film power diode can hold off high voltages (kV's) when operated under reverse bias. The III-nitride device layers can be grown on a wider bandgap template layer and growth substrate, which can be removed by laser lift-off of the epitaxial device layers grown thereon.

  9. Low temperature gaseous nitriding and carburising of stainless steel

    DEFF Research Database (Denmark)

    Christiansen, Thomas; Somers, Marcel A.J.

    2005-01-01

    The response of various austenitic and duplex stainless steel grades to low temperature gaseous nitriding and carburising was investigated. Gaseous nitriding was performed in ammonia/hydrogen mixtures at temperatures ,723 K; gaseous carburising was carried out in carbon monoxide/hydrogen mixtures...

  10. Progress in preparation, properties and application of boron nitride nanomaterials

    Science.gov (United States)

    Wang, Youjun; Han, Jiaqi; Li, Yanjiao; Chen, Hao

    2017-08-01

    Boron nitride nanomaterials have attracted much and more interest in scientific research workers because of their excellent physical and chemical properties. They have become an important research hotspot in today's materials field. In this paper, boron nitride nanoparticles, "fullerenes", nanotubes, nanoribbons and Nano sheets were reviewed in terms of preparation methods, properties and potential applications.

  11. Molybdenum enhanced low-temperature deposition of crystalline silicon nitride

    Science.gov (United States)

    Lowden, Richard A.

    1994-01-01

    A process for chemical vapor deposition of crystalline silicon nitride which comprises the steps of: introducing a mixture of a silicon source, a molybdenum source, a nitrogen source, and a hydrogen source into a vessel containing a suitable substrate; and thermally decomposing the mixture to deposit onto the substrate a coating comprising crystalline silicon nitride containing a dispersion of molybdenum silicide.

  12. Modeling the kinetics of the nitriding and nitrocarburizing of iron

    DEFF Research Database (Denmark)

    Somers, Marcel A. J.; Mittemeijer, Eric J.

    1998-01-01

    The growth kinetics of the iron-nitride compound layer during nitriding and nitrocarburizing of pure iron has been investigated for various temperatures and various combinations of imposed nitrogen and carbon activities. The results indicate that no local equilibrium occurs at the gas/solid inter...

  13. Nitrogen ion irradiation of Au(110) : formation of gold nitride

    NARCIS (Netherlands)

    Šiller, L.; Hunt, M.R.C.; Brown, J.W.; Coquel, J-M.; Rudolf, P.

    2002-01-01

    Often metal nitrides posses unique properties for applications, such as great hardness, high melting points, chemical stability, novel electrical and magnetic properties. One route to the formation of metal nitride films is through ion irradiation of metal surfaces. In this report, the results of ir

  14. Vertical III-nitride thin-film power diode

    Science.gov (United States)

    Wierer, Jr., Jonathan; Fischer, Arthur J.; Allerman, Andrew A.

    2017-03-14

    A vertical III-nitride thin-film power diode can hold off high voltages (kV's) when operated under reverse bias. The III-nitride device layers can be grown on a wider bandgap template layer and growth substrate, which can be removed by laser lift-off of the epitaxial device layers grown thereon.

  15. Continuous and discontinuous precipitation in Fe-1 at.%Cr-1 at.%Mo alloy upon nitriding; crystal structure and composition of ternary nitrides

    Science.gov (United States)

    Steiner, Tobias; Ramudu Meka, Sai; Rheingans, Bastian; Bischoff, Ewald; Waldenmaier, Thomas; Yeli, Guma; Martin, Tomas L.; Bagot, Paul A. J.; Moody, Michael P.; Mittemeijer, Eric J.

    2016-05-01

    The internal nitriding response of a ternary Fe-1 at.%Cr-1 at.%Mo alloy, which serves as a model alloy for many CrMo-based steels, was investigated. The nitrides developing upon nitriding were characterised by X-ray diffraction, scanning electron microscopy, electron probe microanalysis, transmission electron microscopy and atom probe tomography. The developed nitrides were shown to be (metastable) ternary mixed nitrides, which exhibit complex morphological, compositional and structural transformations as a function of nitriding time. Analogous to nitrided binary Fe-Cr and Fe-Mo alloys, in ternary Fe-Cr-Mo alloys initially continuous precipitation of fine, coherent, cubic, NaCl-type nitride platelets, here with the composition (Cr½,Mo½)N¾, occurs, with the broad faces of the platelets parallel to the {1 0 0}α-Fe lattice planes. These nitrides undergo a discontinuous precipitation reaction upon prolonged nitriding leading to the development of lamellae of a novel, hexagonal CrMoN2 nitride along {1 1 0}α-Fe lattice planes, and of spherical cubic, NaCl-type (Cr,Mo)Nx nitride particles within the ferrite lamellae. The observed structural and compositional changes of the ternary nitrides have been attributed to the thermodynamic and kinetic constraints for the internal precipitation of (misfitting) nitrides in the ferrite matrix.

  16. Low Temperature Gaseous Nitriding of a Stainless Steel Containing Strong Nitride Formers

    DEFF Research Database (Denmark)

    Fernandes, Frederico Augusto Pires; Christiansen, Thomas Lundin; Somers, Marcel A. J.

    Low temperature thermochemical surface hardening of the precipitation hardening austenitic stainless steel A286 in solution treated state was investigated. A286 contains, besides high amounts of Cr, also substantial amounts of strong nitride formers as Ti, Al and V. It is shown that simultaneous ...

  17. Epitaxial aluminum nitride tunnel barriers grown by nitridation with a plasma source

    NARCIS (Netherlands)

    Zijlstra, T.; Lodewijk, C.F.J.; Vercruyssen, N.; Tichelaar, F.D.; Loudkov, D.N.; Klapwijk, T.M.

    2007-01-01

    High critical current-density (10 to 420 kA/cm2) superconductor-insulator-superconductor tunnel junctions with aluminum nitride barriers have been realized using a remote nitrogen plasma from an inductively coupled plasma source operated in a pressure range of 10−3–10−1 mbar. We find a much better r

  18. Effect of plasma nitriding and titanium nitride coating on the corrosion resistance of titanium.

    Science.gov (United States)

    Wang, Xianli; Bai, Shizhu; Li, Fang; Li, Dongmei; Zhang, Jing; Tian, Min; Zhang, Qian; Tong, Yu; Zhang, Zichuan; Wang, Guowei; Guo, Tianwen; Ma, Chufan

    2016-09-01

    The passive film on the surface of titanium can be destroyed by immersion in a fluoridated acidic medium. Coating with titanium nitride (TiN) may improve the corrosion resistance of titanium. The purpose of this in vitro study was to investigate the effect of duplex treatment with plasma nitriding and TiN coating on the corrosion resistance of cast titanium. Cast titanium was treated with plasma nitriding and TiN coating. The corrosion resistance of the duplex-treated titanium in fluoride-containing artificial saliva was then investigated through electrochemical and immersion tests. The corroded surface was characterized by scanning electron microscopy (SEM) with energy-dispersive spectroscopy surface scan analysis. The data were analyzed using ANOVA (α=.05) RESULTS: Duplex treatment generated a dense and uniform TiN film with a thickness of 4.5 μm. Compared with untreated titanium, the duplex-treated titanium displayed higher corrosion potential (Ecorr) values (Pplasma nitriding and TiN coating significantly improved the corrosion resistance of cast titanium in a fluoride-containing environment. Copyright © 2016 Editorial Council for the Journal of Prosthetic Dentistry. Published by Elsevier Inc. All rights reserved.

  19. Oxygen Reduction Electrocatalysts Based on Coupled Iron Nitride Nanoparticles with Nitrogen-Doped Carbon

    Directory of Open Access Journals (Sweden)

    Min Jung Park

    2016-06-01

    Full Text Available Aimed at developing a highly active and stable non-precious metal electrocatalyst for oxygen reduction reaction (ORR, a novel FexNy/NC nanocomposite—that is composed of highly dispersed iron nitride nanoparticles supported on nitrogen-doped carbon (NC—was prepared by pyrolyzing carbon black with an iron-containing precursor in an NH3 atmosphere. The influence of the various synthetic parameters such as the Fe precursor, Fe content, pyrolysis temperature and pyrolysis time on ORR performance of the prepared iron nitride nanoparticles was investigated. The formed phases were determined by experimental and simulated X-ray diffraction (XRD of numerous iron nitride species. We found that Fe3N phase creates superactive non-metallic catalytic sites for ORR that are more active than those of the constituents. The optimized Fe3N/NC nanocomposite exhibited excellent ORR activity and a direct four-electron pathway in alkaline solution. Furthermore, the hybrid material showed outstanding catalytic durability in alkaline electrolyte, even after 4,000 potential cycles.

  20. Microwave Study of Field-Effect Devices Based on Graphene/Aluminum Nitride/Graphene Structures

    Science.gov (United States)

    Adabi, Mohammad; Lischner, Johannes; Hanham, Stephen M.; Mihai, Andrei P.; Shaforost, Olena; Wang, Rui; Hao, Ling; Petrov, Peter K.; Klein, Norbert

    2017-03-01

    Metallic gate electrodes are often employed to control the conductivity of graphene based field effect devices. The lack of transparency of such electrodes in many optical applications is a key limiting factor. We demonstrate a working concept of a double layer graphene field effect device that utilizes a thin film of sputtered aluminum nitride as dielectric gate material. For this system, we show that the graphene resistance can be modified by a voltage between the two graphene layers. We study how a second gate voltage applied to the silicon back gate modifies the measured microwave transport data at around 8.7 GHz. As confirmed by numerical simulations based on the Boltzmann equation, this system resembles a parallel circuit of two graphene layers with different intrinsic doping levels. The obtained experimental results indicate that the graphene-aluminum nitride-graphene device concept presents a promising technology platform for terahertz- to- optical devices as well as radio-frequency acoustic devices where piezoelectricity in aluminum nitride can also be exploited.