Lattice vibrational properties of transition metal carbides (TiC, ZrC

Indian Academy of Sciences (India)

Lattice vibrational properties of transition metal carbides (TiC, ZrC and HfC) have been presented by including the effects of free-carrier doping and three-body interactions in the rigid shell model. The short-range overlap repulsion is operative up to the second neighbour ions. An excellent agreement has been obtained ...

Square vortex lattice in p-wave superconductors

International Nuclear Information System (INIS)

Shiraishi, J.

1999-01-01

Making use of the Ginzburg Landau equation for isotropic p-wave superconductors, we construct the single vortex solution in part analytically. The fourfold symmetry breaking term arising from the tetragonal symmetry distortion of the Fermi surface is crucial, since this term indicates a fourfold distortion of the vortex core somewhat similar to the one found in d-wave superconductors. This fourfold distortion of the vortex core in turn favors the square vortex lattice as observed recently by small angle neutron scattering (SANS) experiment from Sr 2 RuO 4 . We find that the hexagonal vortex lattice at H = H c1 transforms into the square one for H = H cr = 0.26 H c2 . On the other hand the SANS data does not reveal such transition. The square vortex covers everywhere studied by the SANS implying H cr is very close to H c1 . Therefore some improvement in the present model is certainly desirable. (orig.)

Survey of post-irradiation examinations made of mixed carbide fuels

International Nuclear Information System (INIS)

Coquerelle, M.

1997-01-01

Post-irradiation examinations on mixed carbide, nitride and carbonitride fuels irradiated in fast flux reactors Rapsodie and DFR were carried out during the seventies and early eighties. In this report, emphasis was put on the fission gas release, cladding carburization and head-end gaseous oxidation process of these fuels, in particular, of mixed carbides. (author). 8 refs, 16 figs, 3 tabs

Mixed Uranium/Refractory Metal Carbide Fuels for High Performance Nuclear Reactors

International Nuclear Information System (INIS)

Knight, Travis; Anghaie, Samim

2002-01-01

Single phase, solid-solution mixed uranium/refractory metal carbides have been proposed as an advanced nuclear fuel for advanced, high-performance reactors. Earlier studies of mixed carbides focused on uranium and either thorium or plutonium as a fuel for fast breeder reactors enabling shorter doubling owing to the greater fissile atom density. However, the mixed uranium/refractory carbides such as (U, Zr, Nb)C have a lower uranium densities but hold significant promise because of their ultra-high melting points (typically greater than 3700 K), improved material compatibility, and high thermal conductivity approaching that of the metal. Various compositions of (U, Zr, Nb)C were processed with 5% and 10% metal mole fraction of uranium. Stoichiometric samples were processed from the constituent carbide powders, while hypo-stoichiometric samples with carbon-to-metal (C/M) ratios of 0.92 were processed from uranium hydride, graphite, and constituent refractory carbide powders. Processing techniques of cold uniaxial pressing, dynamic magnetic compaction, sintering, and hot pressing were investigated to optimize the processing parameters necessary to produce high density (low porosity), single phase, solid-solution mixed carbide nuclear fuels for testing. This investigation was undertaken to evaluate and characterize the performance of these mixed uranium/refractory metal carbides for high performance, ultra-safe nuclear reactor applications. (authors)

Lattice dynamics of α boron and of boron carbide

International Nuclear Information System (INIS)

Vast, N.

1999-01-01

The atomic structure and the lattice dynamics of α boron and of B 4 C boron carbide have been studied by Density Functional Theory (D.F.T.) and Density Functional Perturbation Theory (D.F.P.T.). The bulk moduli of the unit-cell and of the icosahedron have been investigated, and the equation of state at zero temperature has been determined. In α boron, Raman diffusion and infrared absorption have been studied under pressure, and the theoretical and experimental Grueneisen coefficients have been compared. In boron carbide, inspection of the theoretical and experimental vibrational spectra has led to the determination of the atomic structure of B 4 C. Finally, the effects of isotopic disorder have been modeled by an exact method beyond the mean-field approximation, and the effects onto the Raman lines has been investigated. The method has been applied to isotopic alloys of diamond and germanium. (author)

Multi-criteria methodology to design a sodium-cooled carbide-fueled Gen-IV reactor

International Nuclear Information System (INIS)

Stauff, N.

2011-01-01

Compared with earlier plant designs (Phenix, Super-Phenix, EFR), Gen IV Sodium-cooled Fast Reactor requires improved economics while meeting safety and non-proliferation criteria. Mixed Oxide (U-Pu)O 2 fuels are considered as the reference fuels due to their important and satisfactory feedback experience. However, innovative carbide (U-Pu)C fuels can be considered as serious competitors for a prospective SFR fleet since carbide-fueled SFRs can offer another type of optimization which might overtake on some aspects the oxide fuel technology. The goal of this thesis is to reveal the potentials of carbide by designing an optimum carbide-fueled SFR with competitive features and a naturally safe behavior during transients. For a French nuclear fleet, a 1500 MW(e) break-even core is considered. To do so, a multi-physic approach was developed taking into account neutronics, fuel thermo-mechanics and thermal-hydraulic at a pre-design stage. Simplified modeling with the calculation of global neutronic feedback coefficients and a quasi-static evaluation was developed to estimate the behavior of a core during overpower transients, loss of flow and/or loss of heat removal transients. The breakthrough of this approach is to provide the designer with an overall view of the iterative process, emphasizing the well-suited innovations and the most efficient directions that can improve the SFR design project.This methodology was used to design a core that benefits from the favorable features of carbide fuels. The core developed is a large carbide-fueled SFR with high power density, low fissile inventory, break-even capability and forgiving behaviors during the un-scrammed transients studied that should prevent using expensive mitigate systems. However, the core-peak burnup is unlikely to significantly exceed 100 MWd/kg because of the large swelling of the carbide fuel leading to quick pellet-clad mechanical interaction and the low creep capacity of carbide. Moderate linear power fuel

A nodal method of calculating power distributions for LWR-type reactors with square fuel lattices

International Nuclear Information System (INIS)

Hoeglund, Randolph.

1980-06-01

A nodal model is developed for calculating the power distribution in the core of a light water reactor with a square fuel lattice. The reactor core is divided into a number of more or less cubic nodes and a nodal coupling equation, which gives the thermal power density in one node as a function of the power densities in the neighbour nodes, is derived from the neutron diffusion equations for two energy groups. The three-dimensional power distribution can be computed iteratively using this coupling equation, for example following the point Jacobi, the Gauss-Seidel or the point successive overrelaxation scheme. The method has been included as the neutronic model in a reactor core simulation computer code BOREAS, where it is combined with a thermal-hydraulic model in order to make a simultaneous computation of the interdependent power and void distributions in a boiling water reactor possible. Also described in this report are a method for temporary one-dimensional iteration developed in order to accelerate the iterative solution of the problem and the Haling principle which is widely used in the planning of reloading operations for BWR reactors. (author)

Silver diffusion through silicon carbide in microencapsulated nuclear fuels TRISO

International Nuclear Information System (INIS)

Cancino T, F.; Lopez H, E.

2013-10-01

The silver diffusion through silicon carbide is a challenge that has persisted in the development of microencapsulated fuels TRISO (Tri structural Isotropic) for more than four decades. The silver is known as a strong emitter of gamma radiation, for what is able to diffuse through the ceramic coatings of pyrolytic coal and silicon carbide and to be deposited in the heat exchangers. In this work we carry out a recount about the art state in the topic of the diffusion of Ag through silicon carbide in microencapsulated fuels and we propose the role that the complexities in the grain limit can have this problem. (Author)

Evaluation of Codisposal Viability for TH/U Carbide (Fort Saint Vrain HTGR) DOE-Owned Fuel

International Nuclear Information System (INIS)

Radulescu, H.

2001-01-01

There are more than 250 forms of US Department of Energy (DOE)-owned spent nuclear fuel (SNF). Due to the variety of the spent nuclear fuel, the National Spent Nuclear Fuel Program has designated nine representative fuel groups for disposal criticality analyses based on fuel matrix, primary fissile isotope, and enrichment. The Fort Saint Vrain reactor (FSVR) SNF has been designated as the representative fuel for the Th/U carbide fuel group. The FSVR SNF consists of small particles (spheres of the order of 0.5-mm diameter) of thorium carbide or thorium and high-enriched uranium carbide mixture, coated with multiple, thin layers of pyrolytic carbon and silicon carbide, which serve as miniature pressure vessels to contain fission products and the U/Th carbide matrix. The coated particles are bound in a carbonized matrix, which forms fuel rods or ''compacts'' that are loaded into large hexagonal graphite prisms. The graphite prisms (or blocks) are the physical forms that are handled in reactor loading and unloading operations, and which will be loaded into the DOE standardized SNF canisters. The results of the analyses performed will be used to develop waste acceptance criteria. The items that are important to criticality control are identified based on the analysis needs and result sensitivities. Prior to acceptance to fuel from the Th/U carbide fuel group for disposal, the important items for the fuel types that are being considered for disposal under the Th/U carbide fuel group must be demonstrated to satisfy the conditions determined in this report

Evaluation of Codisposal Viability for TH/U Carbide (Fort Saint Vrain HTGR) DOE-Owned Fuel

Energy Technology Data Exchange (ETDEWEB)

H. radulescu

2001-09-28

There are more than 250 forms of US Department of Energy (DOE)-owned spent nuclear fuel (SNF). Due to the variety of the spent nuclear fuel, the National Spent Nuclear Fuel Program has designated nine representative fuel groups for disposal criticality analyses based on fuel matrix, primary fissile isotope, and enrichment. The Fort Saint Vrain reactor (FSVR) SNF has been designated as the representative fuel for the Th/U carbide fuel group. The FSVR SNF consists of small particles (spheres of the order of 0.5-mm diameter) of thorium carbide or thorium and high-enriched uranium carbide mixture, coated with multiple, thin layers of pyrolytic carbon and silicon carbide, which serve as miniature pressure vessels to contain fission products and the U/Th carbide matrix. The coated particles are bound in a carbonized matrix, which forms fuel rods or ''compacts'' that are loaded into large hexagonal graphite prisms. The graphite prisms (or blocks) are the physical forms that are handled in reactor loading and unloading operations, and which will be loaded into the DOE standardized SNF canisters. The results of the analyses performed will be used to develop waste acceptance criteria. The items that are important to criticality control are identified based on the analysis needs and result sensitivities. Prior to acceptance to fuel from the Th/U carbide fuel group for disposal, the important items for the fuel types that are being considered for disposal under the Th/U carbide fuel group must be demonstrated to satisfy the conditions determined in this report.

Dancoff Correction in Square and Hexagonal Lattices

Energy Technology Data Exchange (ETDEWEB)

Carlvik, I

1966-11-15

This report presents the results of a series of calculations of Dancoff corrections for square and hexagonal rod lattices. The tables cover a wide range of volume ratios and moderator cross sections. The results were utilized for checking the approximative formula of Sauer and also the modification of Bonalumi to Sauer's formula. The modified formula calculates the Dancoff correction with an accuracy of 0.01 - 0.02 in cases of practical interest. Calculations have also been performed on square lattices with an empty gap surrounding the rods. The results demonstrate the error involved in treating this kind of geometry by means of homogenizing the gap and the moderator. The calculations were made on the Ferranti Mercury computer of AB Atomenergi before it was closed down. Since then FORTRAN routines for Dancoff corrections have been written, and a subroutine DASQHE is included in the report.

UK irradiation experience relevant to advanced carbide fuel concepts for LMFBR's

International Nuclear Information System (INIS)

Bagley, K.Q.; Batey, W.; Paris, R.; Sloss, W.M.; Snape, G.P.

1977-01-01

Despite discouraging prognoses of fabrication and reprocessing problems, it is recognized that the quest for a carbide fuel pin design which fully exploits the favourable density and thermal conductivity of (U,Pu) monocarbide must be maintained. Studies in aid of carbide fuel development have, therefore, continued in the UK in parallel with those on oxide, albeit at a substantially lower level of effort, and a sufficient body of irradiation experience has been accumulated to allow discrimination of realistic fuel pin designs

Irradiation performance of helium-bonded uranium--plutonium carbide fuel elements

International Nuclear Information System (INIS)

Latimer, T.W.; Petty, R.L.; Kerrisk, J.F.; DeMuth, N.S.; Levine, P.J.; Boltax, A.

1979-01-01

The current irradiation program of helium-bonded uranium--plutonium carbide elements is achieving its original goals. By August 1978, 15 of the original 171 helium-bonded elements had reached their goal burnups including one that had reached the highest burnup of any uranium--plutonium carbide element in the U.S.--12.4 at.%. A total of 66 elements had attained burnups over 8 at.%. Only one cladding breach had been identified at that time. In addition, the systematic and coordinated approach to the current steady-state irradiation tests is yielding much needed information on the behavior of helium-bonded carbide fuel elements that was not available from the screening tests (1965 to 1974). The use of hyperstoichiometric (U,Pu)C containing approx. 10 vol% (U,Pu) 2 C 3 appears to combine lower swelling with only a slightly greater tendency to carburize the cladding than single-phase (U,Pu)C. The selected designs are providing data on the relationship between the experimental parameters of fuel density, fuel-cladding gap size, and cladding type and various fuel-cladding mechanical interaction mechanisms

Asymptotic equivalence of Dancoff factors in cylindrical and square fuel cells

International Nuclear Information System (INIS)

Rodrigues, Leticia Jenisch; Leite, Sergio de Queiroz Bogado; Vilhena, Marco Tullio de

2009-01-01

In its classical formulation, the Dancoff factor for a perfectly absorbing fuel rod is defined as the relative reduction in the incurrent of resonance neutrons into the rod in the presence of neighboring rods, as compared to the incurrent into a single fuel rod immersed in an infinite moderator. Alternatively, this factor can be viewed as the probability that a neutron emerging from the surface of a fuel rod will enter another fuel rod without any collision in the moderator or cladding. For perfectly absorbing fuel these definitions are equivalent. In the last years, several works appeared in literature reporting improvements in the calculation of Dancoff factors, using both the classical and the collision probability definitions. So far, collision probabilities have been determined in the WIMS (Winfrith Improved Multi-group Scheme) code by numerical integration of the third order Bickley functions, for cells with both cylindrical and square outer boundaries. In this work, we step further reporting Dancoff factors for perfectly absorbing (Black) and partially absorbing (Grey) fuel rods calculated by the collision probability method, in cluster cells with square outer boundaries. In order to validate the results, comparisons are made with the equivalent cylindricalized cell in hypothetical test cases. The calculation is performed considering specularly reflecting boundary conditions for the square lattice and diffusive reflecting boundary conditions for the cylindrical geometry. The results show the expected asymptotic behavior of the solution with increasing cell sizes. (author)

Safety research needs for carbide and nitride fueled LMFBR's. Final report

International Nuclear Information System (INIS)

Kastenberg, W.E.

1975-01-01

The results of a study initiated at UCLA during the academic year 1974--1975 to evaluate and review the potential safety related research needs for carbide and nitride fueled LMFBR's are presented. The tasks included the following: (1) Review Core and primary system designs for any significant differences from oxide fueled reactors, (2) Review carbide (and nitride) fuel element irradiation behavior, (3) Review reactor behavior in postulated accidents, (4) Examine analytical methods of accident analysis to identify major gaps in models and data, and (5) Examine post accident heat removal. (TSS)

Irradiation of a 19 pin subassembly with mixed carbide fuel in KNK II

Science.gov (United States)

Geithoff, D.; Mühling, G.; Richter, K.

1992-06-01

The presentation deals with the fabrication, irradiation and nondestructive postirradiation examinations of LMR fuel pins with mixed (U, Pu)-carbide fuels. The mixed carbide fuel was fabricated by the European Institute of Transuranium Elements using various fabrication procedures. Fuel composition varied therefore in a wide range of tolerances with respect to oxygen and phase content and microstructure. The 19 carbide pins were irradiated in the fast neutron flux of the KNK II reactor to a burn-up of about 7 at% without any failure in the centre of a KNK "carrier element" at a maximum linear rating of 800 W/cm. After dismantling in the Hot Cells of KfK nondestructive examinations were carried out comprising dimensional controls, radiography, γ-scanning and eddy-current testing. The results indicate differences in fuel behaviour with respect to composition of the fuel.

Harmonically trapped dipolar fermions in a two-dimensional square lattice

DEFF Research Database (Denmark)

Larsen, Anne-Louise G.; Bruun, Georg

2012-01-01

We consider dipolar fermions in a two-dimensional square lattice and a harmonic trapping potential. The anisotropy of the dipolar interaction combined with the lattice leads to transitions between phases with density order of different symmetries. We show that the attractive part of the dipolar...

A high-temperature, short-duration method of fabricating surrogate fuel microkernels for carbide-based TRISO nuclear fuels

International Nuclear Information System (INIS)

Vasudevamurthy, G.; Radecka, A.; Massey, C.

2015-01-01

High-temperature gas-cooled reactor technology is a frontrunner among generation IV nuclear reactor designs. Among the advanced nuclear fuel forms proposed for these reactors, dispersion-type fuel consisting of microencapsulated uranium di-oxide kernels, popularly known as tri-structural isotropic (TRISO) fuel, has emerged as the fuel form of choice. Generation IV gas-cooled fast reactors offer the benefit of recycling nuclear waste with increased burn-ups in addition to producing the required power and hydrogen. Uranium carbide has shown great potential to replace uranium di-oxide for use in these fast spectrum reactors. Uranium carbide microkernels for fast reactor TRISO fuel have traditionally been fabricated by long-duration carbothermic reduction and sintering of precursor uranium dioxide microkernels produced using sol-gel techniques. These long-duration conversion processes are often plagued by issues such as final product purity and process parameters that are detrimental to minor actinide retention. In this context a relatively simple, high-temperature but relatively quick-rotating electrode arc melting method to fabricate microkernels directly from a feedstock electrode was investigated. The process was demonstrated using surrogate tungsten carbide on account of its easy availability, accessibility and the similarity of its melting point relative to uranium carbide and uranium di-oxide.

A high-temperature, short-duration method of fabricating surrogate fuel microkernels for carbide-based TRISO nuclear fuels

Energy Technology Data Exchange (ETDEWEB)

Vasudevamurthy, G.; Radecka, A.; Massey, C. [Virginia Commonwealth Univ., Richmond, VA (United States). High Temperature Materials Lab.

2015-07-01

High-temperature gas-cooled reactor technology is a frontrunner among generation IV nuclear reactor designs. Among the advanced nuclear fuel forms proposed for these reactors, dispersion-type fuel consisting of microencapsulated uranium di-oxide kernels, popularly known as tri-structural isotropic (TRISO) fuel, has emerged as the fuel form of choice. Generation IV gas-cooled fast reactors offer the benefit of recycling nuclear waste with increased burn-ups in addition to producing the required power and hydrogen. Uranium carbide has shown great potential to replace uranium di-oxide for use in these fast spectrum reactors. Uranium carbide microkernels for fast reactor TRISO fuel have traditionally been fabricated by long-duration carbothermic reduction and sintering of precursor uranium dioxide microkernels produced using sol-gel techniques. These long-duration conversion processes are often plagued by issues such as final product purity and process parameters that are detrimental to minor actinide retention. In this context a relatively simple, high-temperature but relatively quick-rotating electrode arc melting method to fabricate microkernels directly from a feedstock electrode was investigated. The process was demonstrated using surrogate tungsten carbide on account of its easy availability, accessibility and the similarity of its melting point relative to uranium carbide and uranium di-oxide.

Search for the non-canonical Ising spin glass on rewired square lattices

Science.gov (United States)

Surungan, Tasrief

2018-03-01

A spin glass (SG) of non-canonical type is a purely antiferromagnetic (AF) system, exemplified by the AF Ising model on a scale free network (SFN), studied by Bartolozzi et al. [ Phys. Rev. B73, 224419 (2006)]. Frustration in this new type of SG is rendered by topological factor and its randomness is caused by random connectivity. As an SFN corresponds to a large dimensional lattice, finding non-canonical SG in lattice with physical dimension is desireable. However, a regular lattice can not have random connectivity. In order to obtain lattices with random connection and preserving the notion of finite dimension, we costructed rewired lattices. We added some extra bonds randomly connecting each site of a regular lattice to its next-nearest neighbors. Very recently, Surungan et al., studied AF Heisenberg system on rewired square lattice and found no SG behavior [AIP Conf. Proc. 1719, 030006 (2016)]. Due to the importance of discrete symmetry for phase transition, here we study similar structure for the Ising model (Z 2 symmetry). We used Monte Carlo simulation with Replica Exchange algorithm. Two types of structures were studied, firstly, the rewired square lattices with one extra bonds added to each site, and secondly, two bonds added to each site. We calculated the Edwards-Anderson paremeter, the commonly used parameter in searching for SG phase. The non-canonical SG is clearly observed in the rewired square lattice with two extra bonds added.

Radiation stability of proton irradiated zirconium carbide

International Nuclear Information System (INIS)

Yang, Yong; Dickerson, Clayton A.; Allen, Todd R.

2009-01-01

The use of zirconium carbide (ZrC) is being considered for the deep burn (DB)-TRISO fuel as a replacement for the silicon carbide coating. The radiation stability of ZrC was studied using 2.6 MeV protons, across the irradiation temperature range from 600 to 900degC and to doses up to 1.75 dpa. The microstructural characterization shows that the irradiated microstructure is comprised of a high density of nanometer-sized dislocation loops, while no irradiation induced amorphization or voids are observed. The lattice expansion induced by point defects is found to increase as the dose increases for the samples irradiated at 600 and 800degC, while for the 900degC irradiation, a slight lattice contraction is observed. The radiation hardening is also quantified using a micro indentation technique for the temperature and doses studies. (author)

Separation of Nuclear Fuel Surrogates from Silicon Carbide Inert Matrix

International Nuclear Information System (INIS)

Baney, Ronald

2008-01-01

The objective of this project has been to identify a process for separating transuranic species from silicon carbide (SiC). Silicon carbide has become one of the prime candidates for the matrix in inert matrix fuels, (IMF) being designed to reduce plutonium inventories and the long half-lives actinides through transmutation since complete reaction is not practical it become necessary to separate the non-transmuted materials from the silicon carbide matrix for ultimate reprocessing. This work reports a method for that required process

Ternary carbide uranium fuels for advanced reactor design applications

International Nuclear Information System (INIS)

Knight, Travis; Anghaie, Samim

1999-01-01

Solid-solution mixed uranium/refractory metal carbides such as the pseudo-ternary carbide, (U, Zr, Nb)C, hold significant promise for advanced reactor design applications because of their high thermal conductivity and high melting point (typically greater than 3200 K). Additionally, because of their thermochemical stability in a hot-hydrogen environment, pseudo-ternary carbides have been investigated for potential space nuclear power and propulsion applications. However, their stability with regard to sodium and improved resistance to attack by water over uranium carbide portends their usefulness as a fuel for advanced terrestrial reactors. An investigation into processing techniques was conducted in order to produce a series of (U, Zr, Nb)C samples for characterization and testing. Samples with densities ranging from 91% to 95% of theoretical density were produced by cold pressing and sintering the mixed constituent carbides at temperatures as high as 2650 K. (author)

Status of steady-state irradiation testing of mixed-carbide fuel designs

International Nuclear Information System (INIS)

Harry, G.R.

1983-01-01

The steady-state irradiation program of mixed-carbide fuels has demonstrated clearly the ability of carbide fuel pins to attain peak burnup greater than 12 at.% and peak fluences of 1.4 x 10 23 n/cm 2 (E > 0.1 MeV). Helium-bonded fuel pins in 316SS cladding have achieved peak burnups of 20.7 at.% (192 MWd/kg), and no breaches have occurred in pins of this design. Sodium-bonded fuel pins in 316SS cladding have achieved peak burnups of 15.8 at.% (146 MWd/kg). Breaches have occurred in helium-bonded fuel pins in PE-16 cladding (approx. 5 at.% burnup) and in D21 cladding (approx. 4 at.% burnup). Sodium-bonded fuel pins achieved burnups over 11 at.% in PE-16 cladding and over 6 at.% in D9 and D21 cladding

An Evaluation of Criticality Margin by an Application of Parallelogram Lattice Arrangement in the Nuclear Fuel Storage Rack

International Nuclear Information System (INIS)

Kim, Song Hyun; Kim, Hong Chul; Shin, Chang Ho; Kim, Jong Kyung; Kim, Kyo Youn

2010-01-01

The criticality evaluation in the nuclear fuel storage rack is essentially required for the prevention of the criticality accident. The square lattice structure of the storage rack is commonly used because it has a simple structure for the storage of the numerous fuel assemblies as well as the good mechanical strength. For the design of the fuel storage rack, the boron plate is commonly used for the criticality reduction. In this study, an arrangement method with the parallelogram lattice structure is proposed for the reduction of the boron concentration or the rack pitch. The criticality margins by the application of the parallelogram lattice were evaluated with MCNP5 code. From the result, the reduction of the boron concentrated in the borated-Al plate was evaluated

GEN IV: Carbide Fuel Elaboration for the 'Futurix Concepts' experiment

International Nuclear Information System (INIS)

Vaudez, Stephane; Riglet-Martial, Chantal; Paret, Laurent; Abonneau, Eric

2008-01-01

In order to collect information on the behaviour of the future GFR (Gas Fast Reactor) fuel under fast neutron irradiation, an experimental irradiation program, called 'Futurix-concepts' has been launched at the CEA. The considered concept is a composite material made of a fissile fuel embedded in an inert ceramic matrix. Fissile fuel pellets are made of UPuN or UPuC while ceramics are SiC for the carbide fuel and TiN for the nitride fuel. This paper focuses on the description of the carbide composite fabrication. The UPuC pellets are manufactured using a metallurgical powder process. Fabrication and handling of the fuels are carried out in glove boxes under a nitrogen atmosphere. Carbide fuel is synthesized by carbo-thermic reduction under vacuum of a mixture of actinide oxide and graphitic carbon up to 1550 deg. C. After ball milling, the UPuC powder is pressed to create hexagonal or spherical compacts. They are then sintered up to 1750 deg. C in order to obtain a density of 85 % of the theoretical one. The sintered pellets are inserted into an inert and tight capsule of SiC. In order to control the gap between the fuel and the matrix precisely, the pellets are abraded. The inert matrix is then filled with the pellets and the whole system is sealed by a BRASiC R process at high temperature under a helium atmosphere. Fabrication of the sample to be irradiated was done in 2006 and the irradiation began in May 2007 in the Phenix reactor. This presentation will detail and discuss the results obtained during this fabrication phase. (authors)

Development of a Robust Tri-Carbide Fueled Reactor for Multi-Megawatt Space Power and Propulsion Applications

International Nuclear Information System (INIS)

Samim Anghaie; Knight, Travis W.; Plancher, Johann; Gouw, Reza

2004-01-01

An innovative reactor core design based on advanced, mixed carbide fuels was analyzed for nuclear space power applications. Solid solution, mixed carbide fuels such as (U,Zr,Nb)c and (U,Zr, Ta)C offer great promise as an advanced high temperature fuel for space power reactors

Analysis of crystallite size and microdeformation crystal lattice the tungsten carbide milling in mill high energy

International Nuclear Information System (INIS)

Silva, F.T. da; Nunes, M.A.M.; Souza, C.P. de; Gomes, U.U.

2010-01-01

The tungsten carbide (WC) has wide application due to its properties like high melting point, high hardness, wear resistance, oxidation resistance and good electrical conductivity. The microstructural characteristics of the starting powders influences the final properties of the carbide. In this context, the use of nanoparticle powders is an efficient way to improve the final properties of the WC. The high energy milling stands out from other processes to obtain nanometric powders due to constant microstructural changes caused by this process. Therefore, the objective is to undertake an analysis of microstructural characteristics on the crystallite size and microdeformations of the crystal lattice using the technique of X-ray diffraction (XRD) using the Rietveld refinement. The results show an efficiency of the milling process to reduce the crystallite size, leading to a significant deformation in the crystal lattice of WC from 5h milling. (author)

Characterization of fuel swelling in helium-bonded carbide fuel pins

International Nuclear Information System (INIS)

Louie, D.L.Y.

1987-08-01

This work is not only the first attempt at characterizing the swelling of (U,Pu)C fuel pellets, but it also represents the only detailed examinations on carbide fuel swelling at high fuel burnups (4 to 16 at. %). This characterization includes the contributions of fission gases, cracks and solid fission products to fuel swelling. Significantly, the contributions of fission gases and cracks were determined by using the image analysis technique (IAT) which allows researchers to take areal measurements of the irradiated fuel porosity and cracks from the photographs of metallographic fuel samples. However, because areal measurements for varying depths in the fuel pellet could not be obtained, the crack areal measurements could not be converted into volumetric quantities. Consequently, in this situation, an areal fuel swelling analysis was used. The macroscopic fission-gas induced fuel swelling (MAS) caused by fission-gas bubbles and pores > 1 μm was determined using the measured irradiated fuel porosity because the measuring range of IAT is limited to bubbles and pores >1 μm. Conversely, for fuel swelling induced by fission-gas bubbles < 1 μm, the microscopic fission-gas induced fuel swelling (MIS) was estimated using an areal fuel swelling model

High 240Pu FTR/EMC experiments and analysis: Carbide fuel and UO2 blanket subassembly worths

International Nuclear Information System (INIS)

Ombrellaro, P.A.

1977-06-01

Carbide-plutonium fuel and UO 2 blanket subassembly worth measurements performed at ANL in the EMC/LWR were analyzed. Composition exchange worth calculations were performed for: (a) the replacement of high- 240 Pu fuel composition for low- 240 Pu fuel composition and carbide-plutonium fuel composition, successively, in the center subassembly of the core; (b) the replacement of low- 240 Pu fuel composition for carbide--plutonium fuel composition in one outer driver subassembly; and (c) the replacement of the radial reflector composition with UO 2 blanket composition in one subassembly of the radial reflector. The composition exchange worth calculations were performed in two-dimensional x,y geometry, using diffusion theory and perturbation theory. Each method produces about the same calculated-to-experimental bias factors

New edge-centered photonic square lattices with flat bands

Science.gov (United States)

Zhang, Da; Zhang, Yiqi; Zhong, Hua; Li, Changbiao; Zhang, Zhaoyang; Zhang, Yanpeng; Belić, Milivoj R.

2017-07-01

We report a new class of edge-centered photonic square lattices with multiple flat bands, and consider in detail two examples: the Lieb-5 and Lieb-7 lattices. In these lattices, there are 5 and 7 sites in the unit cell and in general, the number is restricted to odd integers. The number of flat bands m in the new Lieb lattices is related to the number of sites N in the unit cell by a simple formula m =(N - 1) / 2. The flat bands reported here are independent of the pseudomagnetic field. The properties of lattices with even and odd number of flat bands are different. We consider the localization of light in such Lieb lattices. If the input beam excites the flat-band mode, it will not diffract during propagation, owing to the strong mode localization. In the Lieb-7 lattice, the beam will also oscillate during propagation and still not diffract. The period of oscillation is determined by the energy difference between the two flat bands. This study provides a new platform for investigating light trapping, photonic topological insulators, and pseudospin-mediated vortex generation.

The compatibility of stainless steels with particles and powders of uranium carbide and low-sulphur UCS fuels

International Nuclear Information System (INIS)

Venter, S.

1978-05-01

Slightly hyperstoichiometric (U,Pu)C is a potential nuclear fuel for fast breeder reactors. The excess carbon above the stoichiometric amount results in a higher carbon activity in the fuel, and carbon is transferred to the stainless steel cladding, resulting in embrittlement of the cladding. It is with this problem of carbon transfer from the fuel to the cladding that this thesis is concerned. For practical reasons, UC and not (U,Pu)C was used as the fuel. The theory of decarburisation of carbide fuel and the carburisation of stainless steel, the facilities constructed for the project at the Atomic Energy Board, and the experimental techniques used, including preparation of the fuels, are discussed. The effect of a number of variables of uranium carbide fuel on its compatibility behaviour with stainless steels was investigated, as well as the effect om microstructure and type of stainless steel (304, 304 L and 316) on the rate of carburisation. These studies can be briefly summarised under the following headings: powder-particle size; surface oxidation of uranium carbide; preparation temperature of uranium carbide; low sulfur UCS fuels; uranium sulfide and the microstructure and type of steel. The author concludes that: the effect of surface oxidation and particle size must be taken into account when evaluating out-of-pile tests; the possible effects of surface oxidation must be taken into account when considering vibro-compacted carbide fuels; there is no advantage in replacing a fraction of the carbon atoms by sulphur atoms in slightly hyperstoichiometric carbide fuels, and the type and thermo-mechanical treatment of the stainless steel used as cladding material in a fuel pin is not important as far as the rate of carburisation by the fuel is concerned

Anisotropic square lattice Potts ferromagnet: renormalization group treatment

International Nuclear Information System (INIS)

Oliveira, P.M.C. de; Tsallis, C.

1981-01-01

The choice of a convenient self-dual cell within a real space renormalization group framework enables a satisfactory treatment of the anisotropic square lattice q-state Potts ferromagnet criticality. The exact critical frontier and dimensionality crossover exponent PHI as well as the expected universality behaviour (renormalization flow sense) are recovered for any linear scaling factor b and all values of q(q - [pt

Cluster evolution and critical cluster sizes for the square and triangular lattice Ising models using lattice animals and Monte Carlo simulations

NARCIS (Netherlands)

Eising, G.; Kooi, B. J.

2012-01-01

Growth and decay of clusters at temperatures below T-c have been studied for a two-dimensional Ising model for both square and triangular lattices using Monte Carlo (MC) simulations and the enumeration of lattice animals. For the lattice animals, all unique cluster configurations with their internal

Failure analysis of carbide fuels under transient overpower (TOP) conditions

International Nuclear Information System (INIS)

Nguyen, D.H.

1980-06-01

The failure of carbide fuels in the Fast Test Reactor (FTR) under Transient Overpower (TOP) conditions has been examined. The Beginning-of-Cycle Four (BOC-4) all-oxide base case, at $.50/sec ramp rate was selected as the reference case. A coupling between the advanced fuel performance code UNCLE-T and HCDA Code MELT-IIIA was necessary for the analysis. UNCLE-T was used to determine cladding failure and fuel preconditioning which served as initial conditions for MELT-III calculations. MELT-IIIA determined the time of molten fuel ejection from fuel pin

Renormalisation-group specific heat of the square lattice Potts ferromagnet

International Nuclear Information System (INIS)

Martin, H.O.; Tsallis, C.

1982-01-01

The free and internal energies and specific heat of the q-state Potts ferromagnet are discussed. A real space renormalisation group approach is presented which recovers a considerable amount of exact particular results for all dimensionalities (hypercubic lattices). The square lattice case is calculated in detail by using self-dual clusters (which provide the exact critical point for all q). Comparison with Onsager results (q=2) is satisfactory; the general tendencies for q different 2 (1 [pt

Design and fuel fabrication processes for the AC-3 mixed-carbide irradiation test

International Nuclear Information System (INIS)

Latimer, T.W.; Chidester, K.M.; Stratton, R.W.; Ledergerber, G.; Ingold, F.

1992-01-01

The AC-3 test was a cooperative U.S./Swiss irradiation test of 91 wire-wrapped helium-bonded U-20% Pu carbide fuel pins irradiated to 8.3 at % peak burnup in the Fast Flux Test Facility. The test consisted of 25 pins that contained spherepac fuel fabricated by the Paul Scherrer Institute (PSI) and 66 pins that contained pelletized fuel fabricated by the Los Alamos National Laboratory. Design of AC-3 by LANL and PSI was begun in 1981, the fuel pins were fabricated from 1983 to 1985, and the test was irradiated from 1986 to 1988. The principal objective of the AC-3 test was to compare the irradiation performance of mixed-carbide fuel pins that contained either pelletized or sphere-pac fuel at prototypic fluence and burnup levels for a fast breeder reactor

Statistical mechanics of directed models of polymers in the square lattice

CERN Document Server

Rensburg, J V

2003-01-01

Directed square lattice models of polymers and vesicles have received considerable attention in the recent mathematical and physical sciences literature. These are idealized geometric directed lattice models introduced to study phase behaviour in polymers, and include Dyck paths, partially directed paths, directed trees and directed vesicles models. Directed models are closely related to models studied in the combinatorics literature (and are often exactly solvable). They are also simplified versions of a number of statistical mechanics models, including the self-avoiding walk, lattice animals and lattice vesicles. The exchange of approaches and ideas between statistical mechanics and combinatorics have considerably advanced the description and understanding of directed lattice models, and this will be explored in this review. The combinatorial nature of directed lattice path models makes a study using generating function approaches most natural. In contrast, the statistical mechanics approach would introduce...

GEN IV: Carbide Fuel Elaboration for the 'Futurix Concepts' experiment

Energy Technology Data Exchange (ETDEWEB)

Vaudez, Stephane; Riglet-Martial, Chantal; Paret, Laurent; Abonneau, Eric [Commissariat a l' Energie Atomique (C.E.A.), Direction de l' Energie Nucleaire, Centre d' Etudes de Cadarache, 13108 Saint Paul lez Durance Cedex (France)

2008-07-01

In order to collect information on the behaviour of the future GFR (Gas Fast Reactor) fuel under fast neutron irradiation, an experimental irradiation program, called 'Futurix-concepts' has been launched at the CEA. The considered concept is a composite material made of a fissile fuel embedded in an inert ceramic matrix. Fissile fuel pellets are made of UPuN or UPuC while ceramics are SiC for the carbide fuel and TiN for the nitride fuel. This paper focuses on the description of the carbide composite fabrication. The UPuC pellets are manufactured using a metallurgical powder process. Fabrication and handling of the fuels are carried out in glove boxes under a nitrogen atmosphere. Carbide fuel is synthesized by carbo-thermic reduction under vacuum of a mixture of actinide oxide and graphitic carbon up to 1550 deg. C. After ball milling, the UPuC powder is pressed to create hexagonal or spherical compacts. They are then sintered up to 1750 deg. C in order to obtain a density of 85 % of the theoretical one. The sintered pellets are inserted into an inert and tight capsule of SiC. In order to control the gap between the fuel and the matrix precisely, the pellets are abraded. The inert matrix is then filled with the pellets and the whole system is sealed by a BRASiC{sup R} process at high temperature under a helium atmosphere. Fabrication of the sample to be irradiated was done in 2006 and the irradiation began in May 2007 in the Phenix reactor. This presentation will detail and discuss the results obtained during this fabrication phase. (authors)

Review of the literature for dry reprocessing oxide, metal, and carbide fuel: The AIROX, RAHYD, and CARBOX pyrochemical processes

Energy Technology Data Exchange (ETDEWEB)

Hoyt, R.C.; Rhee, B.W. [Rockwell International Corp., Canoga Park, CA (United States). Energy Systems Group

1979-09-30

The state of the art of dry processing oxide, carbide, and metal fuel has been determined through an extensive literature review. Dry processing in one of the most proliferation resistant fuel reprocessing technologies available to date, and is one of the few which can be exported to other countries. Feasibility has been established for oxide, carbide, and metal fuel on a laboratory scale, and large-scale experiments on oxide and carbide fuel have shown viability of the dry processing concept. A complete dry processing cycle has been demonstrated by multicycle processing-refabrication-reirradiation experiments on oxide fuel. Additional experimental work is necessary to: (1) demonstrate the complete fuel cycle for carbide and metal fuel, (2) optimize dry processing conditions, and (3) establish fission product behavior. Dry process waste management is easier than for an aqueous processing facility since wastes are primarily solids and gases. Waste treatment can be accomplished by techniques which have been, or are being, developed for aqueous plants.

Analysis of refabricated fuel: determination of carbon in uranium plutonium mixed carbide

International Nuclear Information System (INIS)

Huwyler, S.

1977-09-01

In developing uranium plutonium mixed carbide which represents an advanced fuel for breeder reactors carbon analysis is an important means of determining the stoichiometry. Methods of carbon determination are briefly reviewed. The carbon determination using a LECO WR-12 Carbon Determinator is treated in detail and experience of three years operation communicated. Problems arising from operating the LECO-apparatus in a glove box are discussed. It is pointed out that carbon determination with the LECO-apparatus is a very fast method with good precision and well suited for the routine analysis of mixed carbide fuel. The accuracy of the method is checked by means of a standard. (Auth.)

Correspondence between spanning trees and the Ising model on a square lattice

Science.gov (United States)

Viswanathan, G. M.

2017-06-01

An important problem in statistical physics concerns the fascinating connections between partition functions of lattice models studied in equilibrium statistical mechanics on the one hand and graph theoretical enumeration problems on the other hand. We investigate the nature of the relationship between the number of spanning trees and the partition function of the Ising model on the square lattice. The spanning tree generating function T (z ) gives the spanning tree constant when evaluated at z =1 , while giving the lattice green function when differentiated. It is known that for the infinite square lattice the partition function Z (K ) of the Ising model evaluated at the critical temperature K =Kc is related to T (1 ) . Here we show that this idea in fact generalizes to all real temperatures. We prove that [Z(K ) s e c h 2 K ] 2=k exp[T (k )] , where k =2 tanh(2 K )s e c h (2 K ) . The identical Mahler measure connects the two seemingly disparate quantities T (z ) and Z (K ) . In turn, the Mahler measure is determined by the random walk structure function. Finally, we show that the the above correspondence does not generalize in a straightforward manner to nonplanar lattices.

Calculation of the Flux in a Square Lattice Cell and a Comparison with Measurements

Energy Technology Data Exchange (ETDEWEB)

Apelqvist, G [State Power Board, Stockholm (Sweden)

1961-05-15

A calculation has been made of the thermal neutron flux in a square lattice cell using methods devised by Galanin. The f and L lattice parameters have been expressed in measurable quantities and a comparison made between measured and calculated values.

Extraordinary lateral beaming of sound from a square-lattice phononic crystal

Energy Technology Data Exchange (ETDEWEB)

Bai, Xiaoxue; Qiu, Chunyin; He, Hailong; Peng, Shasha; Ke, Manzhu [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Liu, Zhengyou, E-mail: zyliu@whu.edu.cn [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Institute for Advanced Studies, Wuhan University, Wuhan 430072 (China)

2017-03-03

Highlights: • An extraordinary lateral beaming phenomenon is observed in a finite phononic crystal made of square lattice. • The phenomenon can be explained by the equivalence of the states located around the four corners of the first Brillouin zone. • The lateral beaming behavior enables a simple design of acoustic beam splitters. • In some sense, the phenomenon can be described by a near zero refractive index. - Abstract: This work revisits the sound transmission through a finite phononic crystal of square lattice. In addition to a direct, ordinary transmission through the sample, an extraordinary lateral beaming effect is also observed. The phenomenon stems from the equivalence of the states located around the four corners of the first Brillouin zone. The experimental result agrees well with the theoretical prediction. The lateral beaming behavior enables a simple design for realizing acoustic beam splitters.

Properties of zirconium carbide for nuclear fuel applications

Energy Technology Data Exchange (ETDEWEB)

Katoh, Yutai; Vasudevamurthy, Gokul, E-mail: gvasudev@vcu.edu; Nozawa, Takashi; Snead, Lance L.

2013-10-15

Zirconium carbide (ZrC) is a potential coating, oxygen-gettering, or inert matrix material for advanced high temperature reactor fuels. ZrC has demonstrated attractive properties for these fuel applications including excellent resistance against fission product corrosion and fission product retention capabilities. However, fabrication of ZrC results in a range of stable sub-stoichiometric and carbon-rich compositions with or without substantial microstructural inhomogeneity, textural anisotropy, and a phase separation, leading to variations in physical, chemical, thermal, and mechanical properties. The effects of neutron irradiation at elevated temperatures, currently only poorly understood, are believed to be substantially influenced by those compositional and microstructural features further adding complexity to understanding the key ZrC properties. This article provides a survey of properties data for ZrC, as required by the United States Department of Energy’s advanced fuel programs in support of the current efforts toward fuel performance modeling and providing guidance for future research on ZrC for fuel applications.

Verify Super Double-Heterogeneous Spherical Lattice Model for Equilibrium Fuel Cycle Analysis AND HTR Spherical Super Lattice Model for Equilibrium Fuel Cycle Analysis

International Nuclear Information System (INIS)

Gray S. Chang

2005-01-01

The currently being developed advanced High Temperature gas-cooled Reactors (HTR) is able to achieve a simplification of safety through reliance on innovative features and passive systems. One of the innovative features in these HTRs is reliance on ceramic-coated fuel particles to retain the fission products even under extreme accident conditions. Traditionally, the effect of the random fuel kernel distribution in the fuel pebble/block is addressed through the use of the Dancoff correction factor in the resonance treatment. However, the Dancoff correction factor is a function of burnup and fuel kernel packing factor, which requires that the Dancoff correction factor be updated during Equilibrium Fuel Cycle (EqFC) analysis. An advanced KbK-sph model and whole pebble super lattice model (PSLM), which can address and update the burnup dependent Dancoff effect during the EqFC analysis. The pebble homogeneous lattice model (HLM) is verified by the burnup characteristics with the double-heterogeneous KbK-sph lattice model results. This study summarizes and compares the KbK-sph lattice model and HLM burnup analyzed results. Finally, we discuss the Monte-Carlo coupling with a fuel depletion and buildup code--ORIGEN-2 as a fuel burnup analysis tool and its PSLM calculated results for the HTR EqFC burnup analysis

An optimized BWR fuel lattice for improved fuel utilization

International Nuclear Information System (INIS)

Bernander, O.; Helmersson, S.; Schoen, C.G.

1984-01-01

Optimization of the BWR fuel lattice has evolved into the water cross concept, termed ''SVEA'', whereby the improved moderation within bundles augments reactivity and thus improves fuel cycle economy. The novel design introduces into the assembly a cruciform and double-walled partition containing nonboiling water, thus forming four subchannels, each of which holds a 4x4 fuel rod bundle. In Scandinavian BWRs - for which commercial SVEA reloads are now scheduled - the reactivity gain is well exploited without adverse impact in other respects. In effect, the water cross design improves both mechanical and thermal-hydraulic performance. Increased average burnup is also promoted through achieving flatter local power distributions. The fuel utilization savings are in the order of 10%, depending on the basis of comparison, e.g. choice of discharge burnup and lattice type. This paper reviews the design considerations and the fuel utilization benefits of the water cross fuel for non-Scandinavian BWRs which have somewhat different core design parameters relative to ASEA-ATOM reactors. For one design proposal, comparisons are made with current standard 8x8 fuel rod bundles as well as with 9x9 type fuel in reactors with symmetric or asymmetric inter-assembly water gaps. The effect on reactivity coefficients and shutdown margin are estimated and an assessment is made of thermal-hydraulic properties. Consideration is also given to a novel and advantageous way of including mixed-oxide fuel in BWR reloads. (author)

Square-lattice large-pitch hollow-core photonic crystal fiber

DEFF Research Database (Denmark)

Couny, F.; Roberts, John; Birks, T.A.

2008-01-01

We report on the design, fabrication and characterization of silica square-lattice hollow core photonic crystal fibers optimized for low loss guidance over an extended frequency range in the mid-IR region of the optical spectrum. The fiber's linear optical properties include an ultra-low group...... velocity dispersion and a polarization cross-coupling as low as -13.4dB over 10m of fiber....

RICM, Resonance Absorption in Multi-Region Slab or Square or Hexagonal Lattice

International Nuclear Information System (INIS)

Mizuta, H.; Aoyama, K.; Fukai, Y.

1968-01-01

1 - Nature of physical problem solved: Calculates the resonance absorption integral of resonant isotope in a multi-region lattice using the first flight collision probability. The lattice configurations considered are a slab lattice, a square or hexagonal lattice and a cylindricalized lattice with isotropic or perfect reflecting boundary condition. Cases for an isolated rod or plate and homogeneous system can also be treated. 2 - Method of solution: Slowing down of neutrons by each isotope in each region is solved by either exact numerical integration of the slowing down equation or narrow - or wide-resonance approximation. Breit-Wigner's single level formula is used for the resonance cross section and Porter-Thomas distribution of neutron width is taken into account in the unresolved region. 3 - Restrictions on the complexity of the problem: Maximum number of regions: 5; Maximum Number of groups: 100

Nuclear analysis of the experimental VHTR fuel lattice

International Nuclear Information System (INIS)

Doi, Takeshi; Shindo, Ryuiti; Hirano, Mitsumasa; Takano, Makoto

1984-11-01

Nuclear properties of a fuel lattice in the experimental VHTR core were analyzed with DELIGHT-6 and SRAC codes. Analytical results by both codes were compared by using various calculational model. The nuclear parameters were analyzed, such as a multiplication factor of a fuel lattice and it's variation with burnup, a temperature effect on reactivity, an effect of double-heterogeniety in a resonance absorption calculation, a resonance integral of 238 U and a reactivity worth of burnable poison. From these analyses, following results were obtained. Firstly, on calculational models, 1) Effect of double-heterogeniety in the resonance absorption calculation for Mark-III fuel element, causing a decrease of about 5.5 barns in the resonance integral and an increase of about 2.6 %ΔK in the infinite multiplication factor, 2) The heterogeneous calculation with the collision probability method resulted in about 0.6 %ΔK higher the multiplication factor of fuel lattice than that with the point model, 3) The reactivity worth of burnable poison rod by a multi-region model is about 20 % less than that by a 2-region model at an initial state of burnup and it's variation with burnup are fairly different, Secondly, on comparison between the results by DELIGHT-6 and SRAC, 4) The nuclear parameters obtained with both codes agreed well, Lastly, on the improvement in DELIGHT-6, 5) Consideration of the neutron spectrum shielding effect in the resonance effective cross section calculation caused a decrease of about 2.4 %ΔK in the multiplication factor of fuel lattice, 6) The lattice multiplication factor increased about 0.5 %ΔK by introducing lambda-parameters for the non-resonant nuclie. (J.P.N.)

Assessment of neutron transport codes for application to CANDU fuel lattices analysis

International Nuclear Information System (INIS)

Roh, Gyu Hong; Choi, Hang Bok

1999-08-01

In order to assess the applicability of WIMS-AECL and HELIOS code to the CANDU fuel lattice analysis, the physics calculations has been carried out for the standard CANDU fuel and DUPIC fuel lattices, and the results were compared with those of Monte Carlo code MCNP-4B. In this study, in order to consider the full isotopic composition and the temperature effect, new MCNP libraries have been generated from ENDF/B-VI release 3 and validated for typical benchmark problems. The TRX-1,2,BAPL-1,2,3 pin -cell lattices and KENO criticality safety benchmark calculations have been performed for the new MCNP libraries, and the results have shown that the new MCNP library has sufficient accuracy to be used for physics calculation. Then, the lattice codes have been benchmarked by the MCNP code for the major physics parameters such as the burnup reactivity, void reactivity, relative pin power and Doppler coefficient, etc. for the standard CANDU fuel and DUPIC fuel lattices. For the standard CANDU fuel lattice, it was found that the results of WIMS-AECL calculations are consistent with those of MCNP. For the DUPIC fuel lattice, however, the results of WIMS-AECL calculations with ENDF/B-V library have shown that the discrepancy from the results of MCNP calculations increases when the fuel burnup is relatively high. The burnup reactivities of WIMS-ACEL calculations with ENDF/B-VI library have shown excellent agreements with those of MCNP calculation for both the standard CANDU and DUPIC fuel lattices. However, the Doppler coefficient have relatively large discrepancies compared with MCNP calculations, and the difference increases as the fuel burns. On the other hand, the results of HELIOS calculation are consistent with those of MCNP even though the discrepancy is slightly larger compared with the case of the standard CANDU fuel lattice. this study has shown that the WIMS-AECL products reliable results for the natural uranium fuel. However, it is recommended that the WIMS

Direct calculation of the spin stiffness on square, triangular and cubic lattices using the coupled cluster method

OpenAIRE

Krüger, S. E.; Darradi, R.; Richter, J.; Farnell, D. J. J

2006-01-01

We present a method for the direct calculation of the spin stiffness by means of the coupled cluster method. For the spin-half Heisenberg antiferromagnet on the square, the triangular and the cubic lattices we calculate the stiffness in high orders of approximation. For the square and the cubic lattices our results are in very good agreement with the best results available in the literature. For the triangular lattice our result is more precise than any other result obtained so far by other a...

Comparable studies of magnetic properties of Ising spins-5/2 and 3/2 systems on decorated square and triangular lattices

International Nuclear Information System (INIS)

Masrour, R.; Jabar, A.; Benyoussef, A.; Hamedoun, M.

2016-01-01

In this work, we have studied and compared the magnetic properties of Ising spins-5/2 and 3/2 systems on decorated square and triangular lattices using the Monte Carlo simulations. The transition temperature of the two-dimensional decorated square and triangular lattices has been obtained. The effect of the exchange interactions and crystal field on the magnetization is investigated. The magnetic coercive field and saturation magnetization of the two-dimensional decorated square and triangular lattices have been obtained.

Comparable studies of magnetic properties of Ising spins-5/2 and 3/2 systems on decorated square and triangular lattices

Energy Technology Data Exchange (ETDEWEB)

Masrour, R., E-mail: rachidmasrour@hotmail.com [Laboratory of Materials, Processes, Environment and Quality, Cady Ayyed University, National School of Applied Sciences, 63 46000 Safi (Morocco); Jabar, A. [Laboratory of Materials, Processes, Environment and Quality, Cady Ayyed University, National School of Applied Sciences, 63 46000 Safi (Morocco); Benyoussef, A. [Institute of Nanomaterials and Nanotechnologies, MAScIR, Rabat (Morocco); Hassan II Academy of Science and Technology, Rabat (Morocco); Hamedoun, M. [Institute of Nanomaterials and Nanotechnologies, MAScIR, Rabat (Morocco)

2016-07-15

In this work, we have studied and compared the magnetic properties of Ising spins-5/2 and 3/2 systems on decorated square and triangular lattices using the Monte Carlo simulations. The transition temperature of the two-dimensional decorated square and triangular lattices has been obtained. The effect of the exchange interactions and crystal field on the magnetization is investigated. The magnetic coercive field and saturation magnetization of the two-dimensional decorated square and triangular lattices have been obtained.

Review of experimental studies of zirconium carbide coated fuel particles for high temperature gas-cooled reactors

International Nuclear Information System (INIS)

Minato, Kazuo; Ogawa, Toru; Fukuda, Kousaku

1995-03-01

Experimental studies of zirconium carbide(ZrC) coated fuel particles were reviewed from the viewpoints of fuel particle designs, fabrication, characterization, fuel performance, and fission product retentiveness. ZrC is known as a refractory and chemically stable compound, so ZrC is a candidate to replace the silicon carbide(SiC) coating layer of the Triso-coated fuel particles. The irradiation experiments, the postirradiation heating tests, and the out-of-reactor experiments showed that the ZrC layer was less susceptible than the SiC layer to chemical attack by fission products and fuel kernels, and that the ZrC-coated fuel particles performed better than the standard Triso-coated fuel particles at high temperatures, especially above 1600degC. The ZrC-coated fuel particles demonstrated better cesium retention than the standard Triso-coated fuel particles though the ZrC layer showed a less effective barrier to ruthenium than the SiC layer. (author) 51 refs

Evaluation of temperature coefficients of reactivity for 233U--thorium fueled HTGR lattices. Final report

International Nuclear Information System (INIS)

Newman, D.F.; Leonard, B.R. Jr.; Trapp, T.J.; Gore, B.F.; Kottwitz, D.A.; Thompson, J.K.; Purcell, W.L.; Stewart, K.B.

1977-05-01

A comparison of calculated and measured neutron multiplication factors as a function of temperature was made for three graphite-moderated lattices in the High Temperature Lattice Test Reactor (HTLTR) using 233 UO 2 --ThO 2 fuels in varying amounts and configurations. Correlation of neutronic analysis methods and cross section data with the experimental measurements forms the basis for assessing the accuracy of the methods and data and developing confidence in the ability to predict the temperature coefficient of reactivity for various High Temperature Gas-Cooled Reactor (HTGR) conditions in which 233 U and thorium are present in the fuel. The calculated values of k/sub infinity/(T) were correlated with measured values using two least-squares-fitted correlation coefficients: (1) a normalization factor, and (2) a temperature coefficient bias factor. These correlations indicate the existence of a negative (nonconservative) bias in temperature coefficients of reactivity calculated using ENDF/B-IV cross section data

Fuel lattice design in a boiling water reactor using a knowledge-based automation system

International Nuclear Information System (INIS)

Tung, Wu-Hsiung; Lee, Tien-Tso; Kuo, Weng-Sheng; Yaur, Shung-Jung

2015-01-01

Highlights: • An automation system was developed for the fuel lattice radial design of BWRs. • An enrichment group peaking equalizing method is applied to optimize the design. • Several heuristic rules and restrictions are incorporated to facilitate the design. • The CPU time for the system to design a 10x10 lattice was less than 1.2 h. • The beginning-of-life LPF was improved from 1.319 to 1.272 for one of the cases. - Abstract: A knowledge-based fuel lattice design automation system for BWRs is developed and applied to the design of 10 × 10 fuel lattices. The knowledge implemented in this fuel lattice design automation system includes the determination of gadolinium fuel pin location, the determination of fuel pin enrichment and enrichment distribution. The optimization process starts by determining the gadolinium distribution based on the pin power distribution of a flat enrichment lattice and some heuristic rules. Next, a pin power distribution flattening and an enrichment grouping process are introduced to determine the enrichment of each fuel pin enrichment type and the initial enrichment distribution of a fuel lattice design. Finally, enrichment group peaking equalizing processes are performed to achieve lower lattice peaking. Several fuel lattice design constraints are also incorporated in the automation system such that the system can accomplish a design which meets the requirements of practical use. Depending on the axial position of the lattice, a different method is applied in the design of the fuel lattice. Two typical fuel lattices with U"2"3"5 enrichment of 4.471% and 4.386% were taken as references. Application of the method demonstrates that improved lattice designs can be achieved through the enrichment grouping and the enrichment group peaking equalizing method. It takes about 11 min and 1 h 11 min of CPU time for the automation system to accomplish two design cases on an HP-8000 workstation, including the execution of CASMO-4 lattice

Fuel lattice design in a boiling water reactor using a knowledge-based automation system

Energy Technology Data Exchange (ETDEWEB)

Tung, Wu-Hsiung, E-mail: wstong@iner.gov.tw; Lee, Tien-Tso; Kuo, Weng-Sheng; Yaur, Shung-Jung

2015-11-15

Highlights: • An automation system was developed for the fuel lattice radial design of BWRs. • An enrichment group peaking equalizing method is applied to optimize the design. • Several heuristic rules and restrictions are incorporated to facilitate the design. • The CPU time for the system to design a 10x10 lattice was less than 1.2 h. • The beginning-of-life LPF was improved from 1.319 to 1.272 for one of the cases. - Abstract: A knowledge-based fuel lattice design automation system for BWRs is developed and applied to the design of 10 × 10 fuel lattices. The knowledge implemented in this fuel lattice design automation system includes the determination of gadolinium fuel pin location, the determination of fuel pin enrichment and enrichment distribution. The optimization process starts by determining the gadolinium distribution based on the pin power distribution of a flat enrichment lattice and some heuristic rules. Next, a pin power distribution flattening and an enrichment grouping process are introduced to determine the enrichment of each fuel pin enrichment type and the initial enrichment distribution of a fuel lattice design. Finally, enrichment group peaking equalizing processes are performed to achieve lower lattice peaking. Several fuel lattice design constraints are also incorporated in the automation system such that the system can accomplish a design which meets the requirements of practical use. Depending on the axial position of the lattice, a different method is applied in the design of the fuel lattice. Two typical fuel lattices with U{sup 235} enrichment of 4.471% and 4.386% were taken as references. Application of the method demonstrates that improved lattice designs can be achieved through the enrichment grouping and the enrichment group peaking equalizing method. It takes about 11 min and 1 h 11 min of CPU time for the automation system to accomplish two design cases on an HP-8000 workstation, including the execution of CASMO-4

Solitary heat waves in nonlinear lattices with squared on-site potential

Indian Academy of Sciences (India)

A model Hamiltonian is proposed for heat conduction in a nonlinear lattice with squared on-site potential using the second quantized operators and averaging the same using a suitable wave function, equations are derived in discrete form for the field amplitude and the properties of heat transfer are examined theoretically.

Implementation of the Least-Squares Lattice with Order and Forgetting Factor Estimation for FPGA

Czech Academy of Sciences Publication Activity Database

Pohl, Zdeněk; Tichý, Milan; Kadlec, Jiří

2008-01-01

Roč. 2008, č. 2008 (2008), s. 1-11 ISSN 1687-6172 R&D Projects: GA MŠk(CZ) 1M0567 EU Projects: European Commission(XE) 027611 - AETHER Program:FP6 Institutional research plan: CEZ:AV0Z10750506 Keywords : DSP * Least-squares lattice * order estimation * exponential forgetting factor estimation * FPGA implementation * scheduling * dynamic reconfiguration * microblaze Subject RIV: IN - Informatics, Computer Science Impact factor: 1.055, year: 2008 http://library.utia.cas.cz/separaty/2008/ZS/pohl-tichy-kadlec-implementation%20of%20the%20least-squares%20lattice%20with%20order%20and%20forgetting%20factor%20estimation%20for%20fpga.pdf

Evaluation of catalytic properties of tungsten carbide for the anode of microbial fuel cells

Energy Technology Data Exchange (ETDEWEB)

Rosenbaum, Miriam; Zhao, Feng; Quaas, Marion; Wulff, Harm; Schroeder, Uwe; Scholz, Fritz [Universitaet Greifswald, Institut fuer Biochemie, Felix-Hausdorff-Strasse 4, 17487 Greifswald (Germany)

2007-07-31

In this communication we discuss the properties of tungsten carbide, WC, as anodic electrocatalyst for microbial fuel cell application. The electrocatalytic activity of tungsten carbide is evaluated in the light of its preparation procedure, its structural properties as well as the pH and the composition of the anolyte solution and the catalyst load. The activity of the noble-metal-free electrocatalyst towards the oxidation of several common microbial fermentation products (hydrogen, formate, lactate, ethanol) is studied for microbial fuel cell conditions (e.g., pH 5, room temperature and ambient pressure). Current densities of up to 8.8 mA cm{sup -2} are achieved for hydrogen (hydrogen saturated electrolyte solution), and up to 2 mA cm{sup -2} for formate and lactate, respectively. No activity was observed for ethanol electrooxidation. The electrocatalytic activity and chemical stability of tungsten carbide is excellent in acidic to pH neutral potassium chloride electrolyte solutions, whereas higher phosphate concentrations at neutral pH support an oxidative degradation. (author)

Computer simulation of trails on a square lattice. I. Trails at infinite temperature

International Nuclear Information System (INIS)

Lim, H.A.; Meirovitch, H.

1989-01-01

A trail is a random walk on a lattice for which two bonds are not allowed to overlap. However, the chain may cross itself and one may associate with each such intersection an attractive energy epsilon-c. We study trails at infinite temperature T = ∞ (i.e., trails without attractions) on a square lattice using the scanning simulation method. Our results for the radius of gyration and the end-to-end distance strongly suggest (as do previous studies) that the shape exponent is ν = 0.75, similar to that for self-avoiding walks (SAW's). We obtain significantly more accurate estimates than have been obtained before for the entropy exponent γ = 1.350 +- 0.012 and for the effective growth parameter μ = 2.720 58 +- 0.000 20 (95% confidence limit). The persistence length is found to increase with increasing chain length N and the data fit slightly better an exponential function N/sup w/ where w = 0.047 +- 0.009 than a logarithmic one. Guttmann [J. Phys. A 18, 567 (1985)] has shown exactly that trails and SAW's on the hexagonal lattice at T = ∞ have the same exponents. Our results suggest that this is true also for the square lattice

Quenched bond-dilute Ising ferromagnet in square lattice: thermodynamical properties

International Nuclear Information System (INIS)

Honmura, R.; Sarmento, E.F.; Tsallis, C.

1982-01-01

Within an effective field framework which improves the Molecular Field Approximation, the phase diagram, magnetization, specific heat and susceptibility associated with the quenched bond-dilute Ising ferromagnet in square lattice is calculated. The results are qualitatively (and within certain extent quantitatively) satisfactory; in particular the effects, on the specific heat and susceptibility, of the (eventually) coexisting finite and infinite clusters are exhibited. (Author) [pt

Solitary heat waves in nonlinear lattices with squared on-site potential

Indian Academy of Sciences (India)

Abstract. A model Hamiltonian is proposed for heat conduction in a nonlinear lattice with squared on-site potential using the second quantized operators and averaging the same using a suitable wave function, equations are derived in discrete form for the field amplitude and the prop- erties of heat transfer are examined ...

Post irradiation examinations of uranium-plutonium mixed carbide fuels irradiated at low linear power rate

International Nuclear Information System (INIS)

Maeda, Atsushi; Sasayama, Tatsuo; Iwai, Takashi; Aizawa, Sakuei; Ohwada, Isao; Aizawa, Masao; Ohmichi, Toshihiko; Handa, Muneo

1988-11-01

Two pins containing uranium-plutonium carbide fuels which are different in stoichiometry, i.e. (U,Pu)C 1.0 and (U,Pu)C 1.1 , were constructed into a capsule, ICF-37H, and were irradiated in JRR-2 up to 1.0 at % burnup at the linear heat rate of 420 W/cm. After being cooled for about one year, the irradiated capsule was transferred to the Reactor Fuel Examination Facility where the non-destructive examinations of the fuel pins in the β-γ cells and the destructive ones in two α-γ inert gas atmosphere cells were carried out. The release rates of fission gas were low enough, 0.44 % from (U,Pu)C 1.0 fuel pin and 0.09% from (U,Pu)C 1.1 fuel pin, which is reasonable because of the low central temperature of fuel pellets, about 1000 deg C and is estimated that the release is mainly governed by recoil and knock-out mechanisms. Volume swelling of the fuels was observed to be in the range of 1.3 ∼ 1.6 % for carbide fuels below 1000 deg C. Respective open porosities of (U,Pu)C 1.0 and (U,Pu)C 1.1 fuel were 1.3 % and 0.45 %, being in accordance with the release behavior of fission gas. Metallographic observation of the radial sections of pellets showed the increase of pore size and crystal grain size in the center and middle region of (U,Pu)C 1.0 pellets. The chemical interaction between fuel pellets and claddings in the carbide fuels is the penetration of carbon in the fuels to stainless steel tubes. The depth of corrosion layer in inner sides of cladding tubes ranged 10 ∼ 15 μm in the (U,Pu)C 1.0 fuel and 15 #approx #25 μm in the (U,Pu)C 1.1 fuel, which is correlative with the carbon potential of fuels posibly affecting the amount of carbon penetration. (author)

Optical NOR logic gate design on square lattice photonic crystal platform

Energy Technology Data Exchange (ETDEWEB)

D’souza, Nirmala Maria, E-mail: nirmala@cukerala.ac.in; Mathew, Vincent, E-mail: vincent@cukerala.ac.in [Department of Physics, Central University of Kerala, Kasaragod, Kerala-671 314 (India)

2016-05-06

We numerically demonstrate a new configuration of all-optical NOR logic gate with square lattice photonic crystal (PhC) waveguide using finite difference time domain (FDTD) method. The logic operations are based on interference effect of optical waves. We have determined the operating frequency range by calculating the band structure for a perfectly periodic PhC using plane wave expansion (PWE) method. Response time of this logic gate is 1.98 ps and it can be operated with speed about 513 GB/s. The proposed device consists of four linear waveguides and a square ring resonator waveguides on PhC platform.

RVB signatures in the spin dynamics of the square-lattice Heisenberg antiferromagnet

Science.gov (United States)

Ghioldi, E. A.; Gonzalez, M. G.; Manuel, L. O.; Trumper, A. E.

2016-03-01

We investigate the spin dynamics of the square-lattice spin-\\frac{1}{2} Heisenberg antiferromagnet by means of an improved mean-field Schwinger boson calculation. By identifying both, the long-range Néel and the RVB-like components of the ground state, we propose an educated guess for the mean-field magnetic excitation consisting on a linear combination of local and bond spin flips to compute the dynamical structure factor. Our main result is that when this magnetic excitation is optimized in such a way that the corresponding sum rule is fulfilled, we recover the low- and high-energy spectral weight features of the experimental spectrum. In particular, the anomalous spectral weight depletion at (π,0) found in recent inelastic neutron scattering experiments can be attributed to the interference of the triplet bond excitations of the RVB component of the ground state. We conclude that the Schwinger boson theory seems to be a good candidate to adequately interpret the dynamic properties of the square-lattice Heisenberg antiferromagnet.

Vortex solitons at the interface separating square and hexagonal lattices

Energy Technology Data Exchange (ETDEWEB)

Jović Savić, Dragana, E-mail: jovic@ipb.ac.rs; Piper, Aleksandra; Žikić, Radomir; Timotijević, Dejan

2015-06-19

Vortex solitons at the interface separating two different photonic lattices – square and hexagonal – are demonstrated numerically. We consider the conditions for the existence of discrete vortex states at such interfaces and develop a concise picture of different scenarios of the vortex solutions behavior. Various vortices with different size and topological charges are considered, as well as various lattice interfaces. A novel type of discrete vortex surface solitons in a form of five-lobe solution is observed. Besides stable three-lobe and six-lobe discrete surface modes propagating for long distances, we observe various oscillatory vortex surface solitons, as well as dynamical instabilities of different kinds of solutions and study their angular momentum. Dynamical instabilities occur for higher values of the propagation constant, or at higher beam powers. - Highlights: • We demonstrate vortex solitons at the square–hexagonal photonic lattice interface. • A novel type of five-lobe surface vortex solitons is observed. • Different phase structures of surface solutions are studied. • Orbital angular momentum transfer of such solutions is investigated.

The Heisenberg antiferromagnet on the square-kagomé lattice

Directory of Open Access Journals (Sweden)

J. Richter

2009-01-01

Full Text Available We discuss the ground state, the low-lying excitations as well as high-field thermodynamics of the Heisenberg antiferromagnet on the two-dimensional square-kagomé lattice. This magnetic system belongs to the class of highly frustrated spin systems with an infinite non-trivial degeneracy of the classical ground state as it is also known for the Heisenberg antiferromagnet on the kagomé and on the star lattice. The quantum ground state of the spin-half system is a quantum paramagnet with a finite spin gap and with a large number of non-magnetic excitations within this gap. We also discuss the magnetization versus field curve that shows a plateaux as well as a macroscopic magnetization jump to saturation due to independent localized magnon states. These localized states are highly degenerate and lead to interesting features in the low-temperature thermodynamics at high magnetic fields such as an additional low-temperature peak in the specific heat and an enhanced magnetocaloric effect.

Optimization of BWR fuel lattice enrichment and gadolinia distribution using genetic algorithms and knowledge

International Nuclear Information System (INIS)

Martin-del-Campo, Cecilia; Francois, Juan Luis; Carmona, Roberto; Oropeza, Ivonne P.

2007-01-01

An optimization methodology based on the Genetic Algorithms (GA) method was developed for the design of radial enrichment and gadolinia distributions for boiling water reactor (BWR) fuel lattices. The optimization algorithm was linked to the HELIOS code to evaluate the neutronic parameters included in the objective function. The goal is to search for a fuel lattice with the lowest average enrichment, which satisfy a reactivity target, a local power peaking factor (PPF), lower than a limit value, and an average gadolinia concentration target. The methodology was applied to the design of a 10 x 10 fuel lattice, which can be used in fuel assemblies currently used in the two BWRs operating at Mexico. The optimization process showed an excellent performance because it found forty lattice designs in which the worst one has a better neutronic performance than the reference lattice design. The main contribution of this study is the development of an efficient procedure for BWR fuel lattice design, using GA with an objective function (OF) which saves computing time because it does not require lattice burnup calculations

High burnup, high power irradiation behavior of helium-bonded mixed carbide fuel pins

International Nuclear Information System (INIS)

Levine, P.J.; Nayak, U.P.; Boltax, A.

1983-01-01

Large diameter (9.4 mm) helium-bonded mixed carbide fuel pins were successfully irradiated in EBR-II to high burnup (12%) at high power levels (100 kW/m) with peak cladding midwall temperatures of 550 0 C. The wire-wrapped pins were clad with 0.51-mm-thick, 20% cold-worked Type 316 stainless steel and contained hyperstoichiometric (Usub(0.8)Pusub(0.2))C fuel covering the smeared density range from 75-82% TD. Post-irradiation examinations revealed: extensive fuel-cladding mechanical interaction over the entire length of the fuel column, 35% fission gas release at 12% burnup, cladding carburization and fuel restructuring. (orig.)

Standard and inverse bond percolation of straight rigid rods on square lattices

Science.gov (United States)

Ramirez, L. S.; Centres, P. M.; Ramirez-Pastor, A. J.

2018-04-01

Numerical simulations and finite-size scaling analysis have been carried out to study standard and inverse bond percolation of straight rigid rods on square lattices. In the case of standard percolation, the lattice is initially empty. Then, linear bond k -mers (sets of k linear nearest-neighbor bonds) are randomly and sequentially deposited on the lattice. Jamming coverage pj ,k and percolation threshold pc ,k are determined for a wide range of k (1 ≤k ≤120 ). pj ,k and pc ,k exhibit a decreasing behavior with increasing k , pj ,k →∞=0.7476 (1 ) and pc ,k →∞=0.0033 (9 ) being the limit values for large k -mer sizes. pj ,k is always greater than pc ,k, and consequently, the percolation phase transition occurs for all values of k . In the case of inverse percolation, the process starts with an initial configuration where all lattice bonds are occupied and, given that periodic boundary conditions are used, the opposite sides of the lattice are connected by nearest-neighbor occupied bonds. Then, the system is diluted by randomly removing linear bond k -mers from the lattice. The central idea here is based on finding the maximum concentration of occupied bonds (minimum concentration of empty bonds) for which connectivity disappears. This particular value of concentration is called the inverse percolation threshold pc,k i, and determines a geometrical phase transition in the system. On the other hand, the inverse jamming coverage pj,k i is the coverage of the limit state, in which no more objects can be removed from the lattice due to the absence of linear clusters of nearest-neighbor bonds of appropriate size. It is easy to understand that pj,k i=1 -pj ,k . The obtained results for pc,k i show that the inverse percolation threshold is a decreasing function of k in the range 1 ≤k ≤18 . For k >18 , all jammed configurations are percolating states, and consequently, there is no nonpercolating phase. In other words, the lattice remains connected even when

Radionuclide Inventories for DOE SNF Waste Stream and Uranium/Thorium Carbide Fuels

International Nuclear Information System (INIS)

K.L. Goluoglu

2000-01-01

The objective of this calculation is to generate radionuclide inventories for the Department of Energy (DOE) spent nuclear fuel (SNF) waste stream destined for disposal at the potential repository at Yucca Mountain. The scope of this calculation is limited to the calculation of two radionuclide inventories; one for all uranium/thorium carbide fuels in the waste stream and one for the entire waste stream. These inventories will provide input in future screening calculations to be performed by Performance Assessment to determine important radionuclides

Inspection device for fuel rod restraint by support lattice of fuel assembly

International Nuclear Information System (INIS)

Hasegawa, Isao; Senga, Masatoshi; Kada, Mitoshi.

1991-01-01

An inspection operation section for disposing fuel assembly vertically at predetermined positions, a control section wired therewith, a moving operation section movable in the direction of X, Y and Z axes by a driving signal sent from the control section are disposed to an inspection section main body. A downward bore scope and a upward bore scope, each of such a size as can be inserted to the gaps between the fuel rods, are disposed while opposing to each other for observing the inside of each of cells from above and below in support lattices of fuel assemblies. High performance television cameras are disposed to each of bore scopes to supply images to monitoring televisions in the control section. Thus, a displacing operation section of the inspection operation section is automatically controlled three-dimensionally, the downward bore scope and the upward bore scope are integrally intruded to the inside of the gaps between the predetermined fuel rods from a required height and stopped at a predetermined position, mounted automatically to a required cell of the support lattice to efficiently observe and inspect the fuel rod restraint. (N.H.)

Universal quantum computation with temporal-mode bilayer square lattices

Science.gov (United States)

Alexander, Rafael N.; Yokoyama, Shota; Furusawa, Akira; Menicucci, Nicolas C.

2018-03-01

We propose an experimental design for universal continuous-variable quantum computation that incorporates recent innovations in linear-optics-based continuous-variable cluster state generation and cubic-phase gate teleportation. The first ingredient is a protocol for generating the bilayer-square-lattice cluster state (a universal resource state) with temporal modes of light. With this state, measurement-based implementation of Gaussian unitary gates requires only homodyne detection. Second, we describe a measurement device that implements an adaptive cubic-phase gate, up to a random phase-space displacement. It requires a two-step sequence of homodyne measurements and consumes a (non-Gaussian) cubic-phase state.

Plutonium fuel lattice neutron behavior in inert matrix

International Nuclear Information System (INIS)

Hernandez L, H.; Lucatero, M. A.

2010-10-01

In several countries is had been researching the possibility of using plutonium, as weapon degree as reactor degree, as fuel material in commercial reactors to generate electricity. In special a great development has been in Pressure Water Reactors. However, in Mexico the reactors are Boiling Water Reactors type, reason for which the necessity to considers feasibility to use this fuel type in the reactors of nuclear power plant of Laguna Verde. For this propose a comparison of fuel lattice that compose a fuel assembly is made. The fuel assembly will propose to be used whit in the reactor present different inert matrix, as well as burnable poison. The material that compose the inert matrices used are cerium and zirconia (CeO 2 and ZrO 2 ) and as burnable poisons have gadolinium and erbium (Gd 2 O 4 and ErO 2 ). As far as the hydraulic design used is a cell 10 X 10 with two water channels. The lattice calculations are made with the Helios code a library with 35 energy groups, having determined the pin power factors, the infinite multiplication factor and the neutron flux profiles. (Author)

Gamma scanning of mixed carbide and oxide fuel pins irradiated in FBTR

International Nuclear Information System (INIS)

Jayaraj, V.V.; Padalakshmi, M.; Ulaganathan, T.; Venkiteswaran, C.N.; Divakar, R.; Joseph, Jojo; Bhaduri, A.K.

2016-01-01

Fission in nuclear fuels results in a number of fission products that are gamma emitters in the energy range of 100 keV to 3 MeV. The gamma emitting fission products are therefore amenable for detection by gamma detectors. Assessment of the fission product distribution and their migration behavior through gamma scanning is important for characterizing the in reactor behavior of the fuel. Gamma scanning is an important non destructive technique used to evaluate the behavior of irradiated fuels. As a part of Post Irradiation Examinations (PIE), axial gamma scanning has been carried out on selected fuel pins of the FBTR Mark I mixed carbide fuel sub-assemblies and PFBR MOX test fuel sub-assembly irradiated in FBTR. This paper covers the results of gamma scanning and correlation of gamma scanning results with other PIE techniques

The packing of two species of polygons on the square lattice

International Nuclear Information System (INIS)

Dei Cont, David; Nienhuis, Bernard

2004-01-01

We decorate the square lattice with two species of polygons under the constraint that every lattice edge is covered by only one polygon and every vertex is visited by both types of polygons. We end up with a 24-vertex model which is known in the literature as the fully packed double loop model (FPL 2 ). In the particular case in which the fugacities of the polygons are the same, the model admits an exact solution. The solution is obtained using coordinate Bethe ansatz and provides a closed expression for the free energy. In particular, we find the free energy of the four-colouring model and the double Hamiltonian walk and recover the known entropy of the Ice model. When both fugacities are set equal to 2 the model undergoes an infinite-order phase transition

Determining the minimum required uranium carbide content for HTGR UCO fuel kernels

International Nuclear Information System (INIS)

McMurray, Jacob W.; Lindemer, Terrence B.; Brown, Nicholas R.; Reif, Tyler J.; Morris, Robert N.; Hunn, John D.

2017-01-01

Highlights: • The minimum required uranium carbide content for HTGR UCO fuel kernels is calculated. • More nuclear and chemical factors have been included for more useful predictions. • The effect of transmutation products, like Pu and Np, on the oxygen distribution is included for the first time. - Abstract: Three important failure mechanisms that must be controlled in high-temperature gas-cooled reactor (HTGR) fuel for certain higher burnup applications are SiC layer rupture, SiC corrosion by CO, and coating compromise from kernel migration. All are related to high CO pressures stemming from O release when uranium present as UO 2 fissions and the O is not subsequently bound by other elements. In the HTGR kernel design, CO buildup from excess O is controlled by the inclusion of additional uranium apart from UO 2 in the form of a carbide, UC x and this fuel form is designated UCO. Here general oxygen balance formulas were developed for calculating the minimum UC x content to ensure negligible CO formation for 15.5% enriched UCO taken to 16.1% actinide burnup. Required input data were obtained from CALPHAD (CALculation of PHAse Diagrams) chemical thermodynamic models and the Serpent 2 reactor physics and depletion analysis tool. The results are intended to be more accurate than previous estimates by including more nuclear and chemical factors, in particular the effect of transmuted Pu and Np oxides on the oxygen distribution as the fuel kernel composition evolves with burnup.

A percolation process on the square lattice where large finite clusters are frozen

NARCIS (Netherlands)

van den Berg, J.; de Lima, B.N.B.; Nolin, P.

2012-01-01

In (Aldous, Math. Proc. Cambridge Philos. Soc. 128 (2000), 465-477), Aldous constructed a growth process for the binary tree where clusters freeze as soon as they become infinite. It was pointed out by Benjamini and Schramm that such a process does not exist for the square lattice. This motivated us

Square-lattice random Potts model: criticality and pitchfork bifurcation

International Nuclear Information System (INIS)

Costa, U.M.S.; Tsallis, C.

1983-01-01

Within a real space renormalization group framework based on self-dual clusters, the criticality of the quenched bond-mixed q-state Potts ferromagnet on square lattice is discussed. On qualitative grounds it is exhibited that the crossover from the pure fixed point to the random one occurs, while q increases, through a pitchfork bifurcation; the relationship with Harris criterion is analyzed. On quantitative grounds high precision numerical values are presented for the critical temperatures corresponding to various concentrations of the coupling constants J 1 and J 2 , and various ratios J 1 /J 2 . The pure, random and crossover critical exponents are discussed as well. (Author) [pt

CANDU reactors with reactor grade plutonium/thorium carbide fuel

Energy Technology Data Exchange (ETDEWEB)

Sahin, Suemer [Atilim Univ., Ankara (Turkey). Faculty of Engineering; Khan, Mohammed Javed; Ahmed, Rizwan [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan); Gazi Univ., Ankara (Turkey). Faculty of Technology

2011-08-15

Reactor grade (RG) plutonium, accumulated as nuclear waste of commercial reactors can be re-utilized in CANDU reactors. TRISO type fuel can withstand very high fuel burn ups. On the other hand, carbide fuel would have higher neutronic and thermal performance than oxide fuel. In the present work, RG-PuC/ThC TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 60%. The fuel compacts conform to the dimensions of sintered CANDU fuel compacts are inserted in 37 zircolay rods to build the fuel zone of a bundle. Investigations have been conducted on a conventional CANDU reactor based on GENTILLYII design with 380 fuel bundles in the core. Three mixed fuel composition have been selected for numerical calculation; (1) 10% RG-PuC + 90% ThC; (2) 30% RG-PuC + 70% ThC; (3) 50% RG-PuC + 50% ThC. Initial reactor criticality values for the modes (1), (2) and (3) are calculated as k{sub {infinity}}{sub ,0} = 1.4848, 1.5756 and 1.627, respectively. Corresponding operation lifetimes are {proportional_to} 2.7, 8.4, and 15 years and with burn ups of {proportional_to} 72 000, 222 000 and 366 000 MW.d/tonne, respectively. Higher initial plutonium charge leads to higher burn ups and longer operation periods. In the course of reactor operation, most of the plutonium will be incinerated. At the end of life, remnants of plutonium isotopes would survive; and few amounts of uranium, americium and curium isotopes would be produced. (orig.)

Critical phase for the antiferromagnetic Z(5) model on a square lattice

International Nuclear Information System (INIS)

Baltar, V.L.; Carneiro, G.M.; Pol, M.E.; Zagury, N.

1983-04-01

The existence of a critical phase for the antiferromagnetic Z(5) model on a square lattice is suggested based on results of Monte Carlo (MC) simulations and of Migdal Kadanoff Renormalization Group calculations (MKRG). The MKRG simulates a line of fixed points which it is interpreted as the locus of attraction of a critical phase. The MC simulations are compatible with this interpretation. (Author) [pt

A review of the breeding potentials of carbide, nitride and oxide fueled LMFBRs and GCFRs

International Nuclear Information System (INIS)

Handa, Muneo

1977-11-01

The effects of design parameters in large variation on compound system doubling time of large advanced-fueled LMFBR are described on the base of recent U.S. results. The fuel element design by Combustion Engineering Inc. in step-by-step substitution of the initial oxide fuel subassemblies with carbide ones is explained. Breeding characteristics of the oxide-fueled LMFBR and its potential design modifications are expounded. The gas cooled fast breeder program in West Germany and in the United States are briefed. Definitions of the breeding ratio and doubling time in overall fuel cycle are given. (auth.)

Calculation Of A Lattice Physics Parameter For SBWR Fuel Bundle Design

International Nuclear Information System (INIS)

Sardjono, Y.

1996-01-01

The maximum power peaking factor for Nuclear Power Plant SBWR type is 1.5. The precision for that calculation is related with the result of unit cell analysis each rod in the fuel bundles. This analysis consist of lattice eigenvalue, lattice average diffusion cross section as well as relative power peaking factor in the fuel rod for each fuel bundles. The calculation by using TGBLA computer code which is based on the transport and 168 group diffusion theory. From this calculation can be concluded that the maximum relative power peaking factor is 1.304 and lower than design limit

Real-space renormalization group; application to site percolation in square lattice

International Nuclear Information System (INIS)

Tsallis, C.; Schwachheim, G.

1978-05-01

The real-space renormalization group proposed by Reynolds, Klein and Stanley 1977 to treat the site percolation is analysed and extended . The best among 3 possible definitions of 'percolating' configurations and among 5 possible methods to weight these configurations, are established for percolation in square lattices. The use of n xn square clusters leads, for n = 2 (RKS), n = 3 and n = 4, to √ sub (p) approximately equal to 1.635, √ sub(p) approximately equal to 1.533 and √ sub(p) approximately equal to 1.498, and also to P sub(c) approximately equal to 0.382, P sub(c) approximately equal to 0.388 and P sub(c) approximately equal to 0.398, exhibiting in this way the correct (but slow) tendency towards the best up to date values [pt

Performance of a sphere-pac mixed carbide fuel pin irradiated in the Dounreay Fast Reactor (DFR 527/1 experiment)

International Nuclear Information System (INIS)

Bischoff, K.; Smith, L.; Stratton, R.W.

1980-10-01

The DFR 527/1 experiment was the first irradiation of EIR sphere-pac uranium-plutonium mixed carbide fuel in a fast flux. The experiment has been successfully irradiated to a burn-up of 7.3% FIMA at ratings between 45 and 62 kW m - 1 and clad temperatures between 300 and 600 0 C. Restructuring and elemental redistribution has been found to be similar to the pattern established for pellet type fuel and follows effects seen in earlier sphere-pac carbide tests. Gas release of 12-14% has been measured. A preliminary comparison of radial temperature distribution calculations using a first version of the fuel behaviour modelling code SPECKLE with the actual metallography has been attempted. (Auth.)

Lattice dynamics of {alpha} boron and of boron carbide; Proprietes vibrationnelles du bore {alpha} et du carbure de bore

Energy Technology Data Exchange (ETDEWEB)

Vast, N

1999-07-01

The atomic structure and the lattice dynamics of {alpha} boron and of B{sub 4}C boron carbide have been studied by Density Functional Theory (D.F.T.) and Density Functional Perturbation Theory (D.F.P.T.). The bulk moduli of the unit-cell and of the icosahedron have been investigated, and the equation of state at zero temperature has been determined. In {alpha} boron, Raman diffusion and infrared absorption have been studied under pressure, and the theoretical and experimental Grueneisen coefficients have been compared. In boron carbide, inspection of the theoretical and experimental vibrational spectra has led to the determination of the atomic structure of B{sub 4}C. Finally, the effects of isotopic disorder have been modeled by an exact method beyond the mean-field approximation, and the effects onto the Raman lines has been investigated. The method has been applied to isotopic alloys of diamond and germanium. (author)

Thorium Fuel Performance in a Tight-Pitch Light Water Reactor Lattice

International Nuclear Information System (INIS)

Kim, Taek Kyum; Downar, Thomas J.

2002-01-01

Research on the utilization of thorium-based fuels in the intermediate neutron spectrum of a tight-pitch light water reactor (LWR) lattice is reported. The analysis was performed using the Studsvik/Scandpower lattice physics code HELIOS. The results show that thorium-based fuels in the intermediate spectrum of tight-pitch LWRs have considerable advantages in terms of conversion ratio, reactivity control, nonproliferation characteristics, and a reduced production of long-lived radiotoxic wastes. Because of the high conversion ratio of thorium-based fuels in intermediate spectrum reactors, the total fissile inventory required to achieve a given fuel burnup is only 11 to 17% higher than that of 238 U fertile fuels. However, unlike 238 U fertile fuels, the void reactivity coefficient with thorium-based fuels is negative in an intermediate spectrum reactor. This provides motivation for replacing 238 U with 232 Th in advanced high-conversion intermediate spectrum LWRs, such as the reduced-moderator reactor or the supercritical reactor

Effect of Pu-rich agglomerate in MOX fuel on a lattice calculation

International Nuclear Information System (INIS)

Kawashima, Katsuyuki; Yamamoto, Toru; Namekawa, Masakazu

2007-01-01

The effect of Pu-rich agglomerates in U-Pu mixed oxide (MOX) fuel on a lattice calculation has been demonstrated. The Pu-rich agglomerate parameters are defined based on the measurement data of MIMAS-MOX and the focus is on the highly enriched MOX fuel in accordance with increased burnup resulting in a higher volume fraction of the Pu-rich agglomerates. The lattice calculations with a heterogeneous fuel model and a homogeneous fuel model are performed simulating the PWR 17x17 fuel assembly. The heterogeneous model individually treats the Pu-rich agglomerate and U-Pu matrix, whereas the homogeneous model homogenizes the compositions within the fuel pellet. A continuous-energy Monte Carlo burnup code, MVP-BURN, is used for burnup calculations up to 70 GWd/t. A statistical geometry model is applied in modeling a large number of Pu-rich agglomerates assuming that they are distributed randomly within the MOX fuel pellet. The calculated nuclear characteristics include k-inf, Pu isotopic compositions, power density and burnup of the Pu-rich agglomerates, as well as the pellet-averaged Pu compositions as a function of burnup. It is shown that the effect of Pu-rich agglomerates on the lattice calculation is negligibly small. (author)

Percolation of overlapping squares or cubes on a lattice

International Nuclear Information System (INIS)

Koza, Zbigniew; Kondrat, Grzegorz; Suszczyński, Karol

2014-01-01

Porous media are often modeled as systems of overlapping obstacles, which leads to the problem of two percolation thresholds in such systems, one for the porous matrix and the other for the void space. Here we investigate these percolation thresholds in the model of overlapping squares or cubes of linear size k > 1 randomly distributed on a regular lattice. We find that the percolation threshold of obstacles is a nonmonotonic function of k, whereas the percolation threshold of the void space is well approximated by a function linear in 1/k. We propose a generalization of the excluded volume approximation to discrete systems and use it to investigate the transition between continuous and discrete percolation, finding a remarkable agreement between the theory and numerical results. We argue that the continuous percolation threshold of aligned squares on a plane is the same for the solid and void phases and estimate the continuous percolation threshold of the void space around aligned cubes in a 3D space as 0.036(1). We also discuss the connection of the model to the standard site percolation with complex neighborhood. (paper)

Symmetry Breaking Ground States of Bose-Einstein Condensates in 1D Double Square Well and Optical Lattice Well

International Nuclear Information System (INIS)

Yuan Qingxin; Ding Guohui

2005-01-01

We investigate the phenomena of symmetry breaking and phase transition in the ground state of Bose-Einstein condensates (BECs) trapped in a double square well and in an optical lattice well, respectively. By using standing-wave expansion method, we present symmetric and asymmetric ground state solutions of nonlinear Schroedinger equation (NLSE) with a symmetric double square well potential for attractive nonlinearity. In particular, we study the ground state wave function's properties by changing the depth of potential and atomic interactions (here we restrict ourselves to the attractive regime). By using the Fourier grid Hamiltonian method, we also reveal a phase transition of BECs trapped in one-dimensional optical lattice potential.

Neutron multipilication factors as a function of temperature: a comparison of calculated and measured values for lattices using 233UO2-ThO2 fuel in graphite

International Nuclear Information System (INIS)

Newman, D.F.; Gore, B.F.

1978-01-01

Neutron multiplication factors calculated as a function of temperature for three graphite-moderated 233 UO 2 -ThO 2 -fueled lattices are correlated with the values measured for these lattices in the high-temperature lattice test reactor (HTLTR). The correlation analysis is accomplished by fitting calculated values of k/sub infinity/(T) to the measured values using two least-squares-fitted correlation coefficients: (a) a normalization factor and (b) a temperature coefficient bias factor. These correlations indicate the existence of a negative (nonconservative) bias in temperature coefficients of reactivity calculated using ENDF/B-IV cross-section data. Use of an alternate cross-section data set for thorium, which has a smaller resonance integral than ENDF/B-IV data, improved the agreement between calculated and measured temperature coefficients of reactivity for the three experimental lattices. The results of the correlations are used to estimate the bias in the temperature coefficient of reactivity calculated for a lattice typical of fresh 233 U recycle fuel for a high-temperature gas-cooled reactor (HTGR). This extrapolation to a lattice having a heavier fissile loading than the experimental lattices is accomplished using a sensitivity analysis of the estimated bias to alternate thorium cross-section data used in calculations of k/sub infinity/(T). The envelope of uncertainty expected to contain the actual values for the temperature coefficient of the reactivity for the 233 U-fueled HTGR lattice studied remains negative at 1600 K (1327 0 C). Although a broader base of experimental data with improved accuracy is always desirable, the existing data base provided by the HTLTR experiments is judged to be adequate for the verification of neutronic calculations for the HTGR containing 233 U fuel at its current state of development

Systematic construction of spin liquids on the square lattice from tensor networks with SU(2) symmetry

Science.gov (United States)

Mambrini, Matthieu; Orús, Román; Poilblanc, Didier

2016-11-01

We elaborate a simple classification scheme of all rank-5 SU(2) spin rotational symmetric tensors according to (i) the onsite physical spin S , (ii) the local Hilbert space V⊗4 of the four virtual (composite) spins attached to each site, and (iii) the irreducible representations of the C4 v point group of the square lattice. We apply our scheme to draw a complete list of all SU(2)-symmetric translationally and rotationally invariant projected entangled pair states (PEPS) with bond dimension D ≤6 . All known SU(2)-symmetric PEPS on the square lattice are recovered and simple generalizations are provided in some cases. More generally, to each of our symmetry class can be associated a (D -1 )-dimensional manifold of spin liquids (potentially) preserving lattice symmetries and defined in terms of D -independent tensors of a given bond dimension D . In addition, generic (low-dimensional) families of PEPS explicitly breaking either (i) particular point-group lattice symmetries (lattice nematics) or (ii) time-reversal symmetry (chiral spin liquids) or (iii) SU(2) spin rotation symmetry down to U(1 ) (spin nematics or Néel antiferromagnets) can also be constructed. We apply this framework to search for new topological chiral spin liquids characterized by well-defined chiral edge modes, as revealed by their entanglement spectrum. In particular, we show how the symmetrization of a double-layer PEPS leads to a chiral topological state with a gapless edge described by a SU (2) 2 Wess-Zumino-Witten model.

Statistical mechanics of directed models of polymers in the square lattice

International Nuclear Information System (INIS)

Rensburg, E J Janse van

2003-01-01

Directed square lattice models of polymers and vesicles have received considerable attention in the recent mathematical and physical sciences literature. These are idealized geometric directed lattice models introduced to study phase behaviour in polymers, and include Dyck paths, partially directed paths, directed trees and directed vesicles models. Directed models are closely related to models studied in the combinatorics literature (and are often exactly solvable). They are also simplified versions of a number of statistical mechanics models, including the self-avoiding walk, lattice animals and lattice vesicles. The exchange of approaches and ideas between statistical mechanics and combinatorics have considerably advanced the description and understanding of directed lattice models, and this will be explored in this review. The combinatorial nature of directed lattice path models makes a study using generating function approaches most natural. In contrast, the statistical mechanics approach would introduce partition functions and free energies, and then investigate these using the general framework of critical phenomena. Generating function and statistical mechanics approaches are closely related. For example, questions regarding the limiting free energy may be approached by considering the radius of convergence of a generating function, and the scaling properties of thermodynamic quantities are related to the asymptotic properties of the generating function. In this review the methods for obtaining generating functions and determining free energies in directed lattice path models of linear polymers is presented. These methods include decomposition methods leading to functional recursions, as well as the Temperley method (that is implemented by creating a combinatorial object, one slice at a time). A constant term formulation of the generating function will also be reviewed. The thermodynamic features and critical behaviour in models of directed paths may be

Exact low-temperature series expansion for the partition function of the zero-field Ising model on the infinite square lattice

Science.gov (United States)

Siudem, Grzegorz; Fronczak, Agata; Fronczak, Piotr

2016-01-01

In this paper, we provide the exact expression for the coefficients in the low-temperature series expansion of the partition function of the two-dimensional Ising model on the infinite square lattice. This is equivalent to exact determination of the number of spin configurations at a given energy. With these coefficients, we show that the ferromagnetic–to–paramagnetic phase transition in the square lattice Ising model can be explained through equivalence between the model and the perfect gas of energy clusters model, in which the passage through the critical point is related to the complete change in the thermodynamic preferences on the size of clusters. The combinatorial approach reported in this article is very general and can be easily applied to other lattice models. PMID:27721435

Fuel fabrication processes, design and experimental conditions for the joint US-Swiss mixed carbide test in FFTF (AC-3 test)

International Nuclear Information System (INIS)

Stratton, R.W.; Ledergerber, G.; Ingold, F.; Latimer, T.W.; Chidester, K.M.

1993-01-01

The preparation of mixed carbide fuel for a joint (US-Swiss) irradiation test in the US Fast Flux Test Facility (FFTF) is described, together with the experiment design and the irradiation conditions. Two fabrication routes were compared. The US produced 66 fuel pins containing pellet fuel via the powder-pellet (dry) route, and the Swiss group produced 25 sphere pac pins of mixed carbide using the internal gelation (wet) route. Both sets of fuel met all t the requirements of the specifications concerning soichiometry, chemical composition and structure. The pin designs were as similar as possible. The test operated successfully in the FFTF for 620 effective full power days until October 1988 and reached over 8% burn up with peak powers of around 80 kW/m. The conclusions were that the choice of sphere pac or pellet fuel for reactor application is dependent on preferred differences in fabrication (e.g. economics and environmental factors) and not on differences in irradiation behaviour. (orig.)

Conductivity of a square-lattice bond-mixed resistor network

International Nuclear Information System (INIS)

Costa, U.M.S.; Tsallis, C.; Schwaccheim, G.

1985-01-01

Within a real-space renormalization-group framework based on self-dual clusters, the conductivity of a square-lattice quenched bond-random resistor network is calculated, the conductance on each bond being g 1 or g 2 with probabilities (1-p) and p respectively. The group recovers several already known exact results (including slopes), and is consequently believed to be numerically quite reliable for almost all values of p, and all ratios g 1 /g 2 (in particular, g 1 =0 and g 1 =infinite with finite g 2 respectively correspond to the insulator-resitor and superconductor-resistor mixtures). In addition to that, an heuristic analytic expression is proposed for the conductivity which is believed to be a quite satisfactory approximation everywhere not too close to the percolation point. (Author) [pt

Angle-resolved spin wave band diagrams of square antidot lattices studied by Brillouin light scattering

Energy Technology Data Exchange (ETDEWEB)

Gubbiotti, G.; Tacchi, S. [Istituto Officina dei Materiali del Consiglio Nazionale delle Ricerche (IOM-CNR), Sede di Perugia, c/o Dipartimento di Fisica e Geologia, Via A. Pascoli, I-06123 Perugia (Italy); Montoncello, F.; Giovannini, L. [Dipartimento di Fisica e Scienze della Terra, Università di Ferrara, Via G. Saragat 1, I-44122 Ferrara (Italy); Madami, M.; Carlotti, G. [Dipartimento di Fisica e Geologia, Università di Perugia, Via A. Pascoli, I-06123 Perugia (Italy); Ding, J.; Adeyeye, A. O. [Information Storage Materials Laboratory, Department of Electrical and Computer Engineering, National University of Singapore, Singapore 117576 (Singapore)

2015-06-29

The Brillouin light scattering technique has been exploited to study the angle-resolved spin wave band diagrams of squared Permalloy antidot lattice. Frequency dispersion of spin waves has been measured for a set of fixed wave vector magnitudes, while varying the wave vector in-plane orientation with respect to the applied magnetic field. The magnonic band gap between the two most dispersive modes exhibits a minimum value at an angular position, which exclusively depends on the product between the selected wave vector magnitude and the lattice constant of the array. The experimental data are in very good agreement with predictions obtained by dynamical matrix method calculations. The presented results are relevant for magnonic devices where the antidot lattice, acting as a diffraction grating, is exploited to achieve multidirectional spin wave emission.

Thermodynamic properties of magnetic strings on a square lattice

Science.gov (United States)

Mol, Lucas; Oliveira, Denis Da Mata; Bachmann, Michael

2015-03-01

In the last years, spin ice systems have increasingly attracted attention by the scientific community, mainly due to the appearance of collective excitations that behave as magnetic monopole like particles. In these systems, geometrical frustration induces the appearance of degenerated ground states characterized by a local energy minimization rule, the ice rule. Violations of this rule were shown to behave like magnetic monopoles connected by a string of dipoles that carries the magnetic flux from one monopole to the other. In order to obtain a deeper knowledge about the behavior of these excitations we study the thermodynamics of a kind of magnetic polymer formed by a chain of magnetic dipoles in a square lattice. This system is expected to capture the main properties of monopole-string excitations in the artificial square spin ice. It has been found recently that in this geometry the monopoles are confined, but the effective string tension is reduced by entropic effects. To obtain the thermodynamic properties of the strings we have exactly enumerated all possible string configurations of a given length and used standard statistical mechanics analysis to calculate thermodynamic quantities. We show that the low-temperature behavior is governed by strings that satisfy ice rules. Financial support from FAPEMIG and CNPq (Brazilian agencies) are gratefully acknowledged.

Dynamical phase transition in a fully frustrated Josephson array on a square lattice

International Nuclear Information System (INIS)

Fisher, K. D.; Stroud, D.; Janin, L.

1999-01-01

We study dynamical phase transitions at temperature T=0 in a fully frustrated square Josephson junction array subject to a driving current density, which has nonzero components i x , i y parallel to both axes of the lattice. Our numerical results show clear evidence for three dynamical phases: a pinned vortex lattice characterized by zero time-averaged voltages x > t and y > t , a ''plastic'' phase in which both x > t and y > t are nonzero, and a moving lattice phase in which only one of the time-average voltage components is nonzero. The last of these has a finite transverse critical current: if a current is applied in the x direction, a nonzero transverse current density i y is required before y > t becomes nonzero. The voltage traces in the moving lattice phase are periodic in time. By contrast, the voltages in the plastic phase have continuous power spectra that are weakly dependent on frequency. This phase diagram is found numerically to be qualitatively unchanged by the presence of weak disorder. We also describe two simple analytical models that recover some, but not all, the characteristics of the three dynamical phases, and of the phase diagram calculated numerically. (c) 1999 The American Physical Society

Plasmon excitations in doped square-lattice atomic clusters

Science.gov (United States)

Wang, Yaxin; Yu, Ya-Bin

2017-12-01

Employing the tight-binding model, we theoretically study the properties of the plasmon excitations in doped square-lattice atomic clusters. The results show that the dopant atoms would blur the absorption spectra, and give rise to extra plasmon resonant peaks as reported in the literature; however, our calculated external-field induced oscillating charge density shows that no obvious evidences indicate the so-called local mode of plasmon appearing in two-dimensional-doped atomic clusters, but the dopants may change the symmetry of the charge distribution. Furthermore, we show that the disorder of the energy level due to dopant makes the absorption spectrum has a red- or blue-shift, which depends on the position of impurities; disorder of hopping due to dopant makes a blue- or red-shift, a larger (smaller) hopping gives a blue-shift (red-shift); and a larger (smaller) host-dopant and dopant-dopant intersite coulomb repulsion induces a blue-shift (red-shift).

Minimizing the power peaking factor of fuel lattices using factors of group for boiling water reactors

International Nuclear Information System (INIS)

Guzman, J. R.; Longoria, L. C.; De la Cruz, E.; Arredondo, C.

2010-10-01

A method to design the distribution and composition of nuclear fuel for the array of fuel rods in a lattice for BWRs is presented in this work. The aim of the method is to minimize the power peaking factor until an adequate value is reached. Also, this method uses a few calculations of lattice. The method is based on the classification of the fuel rods in two groups: the group of fuel rods with the higher power level (group pow ), and the other group of fuel rods (no-group pow ). The enrichment of 235 U of each fuel rod of the group pow is multiplied by a factor called group fissile factor (f group ), and the enrichment of 235 U of each fuel rod of the no-group pow is multiplied by a factor called no-group fissile factor (f no-group ). These factors are fitted so that the power peaking factor is minimized. The importance of the method with the use of these two factors is applied to the design of a fuel lattice for BWRs as the Laguna Verde nuclear power plant. The calculations of lattice are made by means of the Helios code. (Author)

Silver diffusion through silicon carbide in microencapsulated nuclear fuels TRISO; Difusion de plata a traves de carburo de silicio en combustibles nucleares microencapsulados TRISO

Energy Technology Data Exchange (ETDEWEB)

Cancino T, F.; Lopez H, E., E-mail: Felix.cancino@cinvestav.edu.mx [IPN, Centro de Investigacion y de Estudios Avanzados, Unidad Saltillo, Av. Industria Metalurgica No. 1062, Col. Ramos Arizpe, 25900 Saltillo, Coahuila (Mexico)

2013-10-15

The silver diffusion through silicon carbide is a challenge that has persisted in the development of microencapsulated fuels TRISO (Tri structural Isotropic) for more than four decades. The silver is known as a strong emitter of gamma radiation, for what is able to diffuse through the ceramic coatings of pyrolytic coal and silicon carbide and to be deposited in the heat exchangers. In this work we carry out a recount about the art state in the topic of the diffusion of Ag through silicon carbide in microencapsulated fuels and we propose the role that the complexities in the grain limit can have this problem. (Author)

Square-lattice magnetism of diaboleite Pb2Cu(OH)4Cl2

Science.gov (United States)

Tsirlin, Alexander A.; Janson, Oleg; Lebernegg, Stefan; Rosner, Helge

2013-02-01

We report on the quasi-two-dimensional magnetism of the natural mineral diaboleite Pb2Cu(OH)4Cl2 with a tetragonal crystal structure, which is closely related to that of the frustrated spin-(1)/(2) magnet PbVO3. Magnetic susceptibility of diaboleite is well described by a Heisenberg spin model on a diluted square lattice with the nearest-neighbor exchange of J≃35 K and about 5% of nonmagnetic impurities. The dilution of the spin lattice reflects the formation of Cu vacancies that are tolerated by the crystal structure of diaboleite. The weak coupling between the magnetic planes triggers the long-range antiferromagnetic order below TN≃11 K. No evidence of magnetic frustration is found. We also analyze the signatures of the long-range order in heat-capacity data, and discuss the capability of identifying magnetic transitions with heat-capacity measurements.

Device for fracturing silicon-carbide coatings on nuclear-fuel particles

Science.gov (United States)

Turner, L.J.; Willey, M.G.; Tiegs, S.M.; Van Cleve, J.E. Jr.

This invention is a device for fracturing particles. It is designed especially for use in hot cells designed for the handling of radioactive materials. In a typical application, the device is used to fracture a hard silicon-carbide coating present on carbon-matrix microspheres containing nuclear-fuel materials, such as uranium or thorium compounds. To promote remote control and facilitate maintenance, the particle breaker is pneumatically operated and contains no moving parts. It includes means for serially entraining the entrained particles on an anvil housed in a leak-tight chamber. The flow rate of the gas is at a value effecting fracture of the particles; preferably, it is at a value fracturing them into product particulates of fluidizable size. The chamber is provided with an outlet passage whose cross-sectional area decreases in the direction away from the chamber. The outlet is connected tangentially to a vertically oriented vortex-flow separator for recovering the product particulates entrained in the gas outflow from the chamber. The invention can be used on a batch or continuous basis to fracture the silicon-carbide coatings on virtually all of the particles fed thereto.

Method for fracturing silicon-carbide coatings on nuclear-fuel particles

Science.gov (United States)

Turner, Lloyd J.; Willey, Melvin G.; Tiegs, Sue M.; Van Cleve, Jr., John E.

1982-01-01

This invention is a device for fracturing particles. It is designed especially for use in "hot cells" designed for the handling of radioactive materials. In a typical application, the device is used to fracture a hard silicon-carbide coating present on carbon-matrix microspheres containing nuclear-fuel material, such as uranium or thorium compounds. To promote remote control and facilitate maintenance, the particle breaker is pneumatically operated and contains no moving parts. It includes means for serially entraining the entrained particles on an anvil housed in a leak-tight chamber. The flow rate of the gas is at a value effecting fracture of the particles; preferably, it is at a value fracturing them into product particulates of fluidizable size. The chamber is provided with an outlet passage whose cross-sectional area decreases in the direction away from the chamber. The outlet is connected tangentially to a vertically oriented vortex-flow separator for recovering the product particulates entrained in the gas outflow from the chamber. The invention can be used on a batch or continuous basis to fracture the silicon-carbide coatings on virtually all of the particles fed thereto.

Effect of composition and heat treatment on carbide phases in Ni-Mo alloys

International Nuclear Information System (INIS)

Svistunova, T.V.; Tsvigunov, A.N.; Stegnukhina, L.V.; Sakuta, N.D.

1984-01-01

The investigation results of vanadium, iron, carbon and silicon effect and heat treatment regime on the type and composition of carbides in Ni-(26...31)%Mo alloys are presented. It is shown that type, composition and quantity of carbide phases forming in alloys are determined not only by molybdenum and carbon content, but presence of other elements (V, Fe), admixtures (C, Si) and reducers as well as by regime of thermal treatment. In the alloy, containing 26...31% Mo, 0.01...0.03% C ( 12 C type with a=1.083...1.089 nm lattice parameter, in which V and Ti, Fe and Si are presented besides Mo and Ni. In the temperature range of 600-800 deg C high dispersed carbides segregate on grain boundaries. Silicon initiates segregation of the carbide phases among them by grain boundaries at the temperatures of 800 deg C as well as regulates carbide of M 12 C type with a=1.094...1.098 nm lattice parameter

Structure and thermal expansion of NbC complex carbides

International Nuclear Information System (INIS)

Khatsinskaya, I.M.; Chaporova, I.N.; Cheburaeva, R.F.; Samojlov, A.I.; Logunov, A.V.; Ignatova, I.A.; Dodonova, L.P.

1983-01-01

Alloying dependences of the crystal lattice parameters at indoor temperature and coefficient of thermal linear exspansion within a 373-1273 K range are determined for complex NbC-base carbides by the method of mathematical expemental design. It is shown that temperature changes in the linear expansion coefficient of certain complex carbides as distinct from NbC have an anomaly (minimum) within 773-973 K caused by occurring reversible phase transformations. An increase in the coefficient of thermal linear expansion and a decrease in hardness of NbC-base tungsten-, molybdenum-, vanadium- and hafnium-alloyed carbides show a weakening of a total chemical bond in the complex carbides during alloying

Advanced Silicon Carbide from Molecular Engineering and Actinide Fuels

International Nuclear Information System (INIS)

Meyer, D.J.M.; Garcia, J.; Guillaneux, D.; Wong-Chi-Man, M.; Moreau, J.J.E.

2008-01-01

In the frame of nuclear fuels studies for generation IV, carbides or oxycarbides assemblies are one of the engaged material for high temperature reactors. The design of the fuels is not yet defined but some structures are actually considered with SiC as matrix for the actinide fuel. In this work we have studied the synthesis of a multi-scale structure controlled SiC matrix using molecular silicon organometallic precursors. The aim of this work was to develop a way to obtain multi-scale SiC matrix material which could be engineered to fit in any fuel structure defined for generation IV fuels. The control of this multi-scale structure was done using several simulation methods specific of the low temperature solution synthesis of the precursor. In a first step, we have focused our effort on the synthesis of the SiC material. A first level of template was successfully done by the use of solid silica 500 nm balls. A second level of template was studied by the use of meso-porous silica, structured at a 50 nm level. At least, supra-molecular simulation in non aqueous media was considered with the difficulty to build a molecular assembly (inverse micelles). In a second step, we have functionalized the primary silane phase with actinide complexing agent in order to blend directly the actinide inside this primary phase in a controlled way. During these studies, a new one pot synthesis route to obtain the functionalized primary silane phase was developed. (authors)

A renormalized -group attempt to obtain the exact transition line of the square - lattice bond - dilute Ising model

International Nuclear Information System (INIS)

Tsallis, C.; Levy, S.V.F.

1979-05-01

Two different renormalization-group approaches are used to determine approximate solutions for the paramagnetic-ferromagnetic transition line of the square-lattice bond-dilute first-neighbour-interaction Ising model. (Author) [pt

Influence of nonstoichiometry and ordering on basic structure parameter of cubic titanium carbide

International Nuclear Information System (INIS)

Zueva, L.V.; Gusev, A.I.

1999-01-01

Effect of nonstoichiometry and phase transformations of the disorder-order type on the basis (B1 type) structure period of TiC y (0.5 y titanium carbide with formation of the Ti 2 C and Ti 3 C 2 superstructures leads to growth of the basic crystal lattice period as compared to disordered carbide. The problem on trends in static atomic displacement near vacancy is discussed with an account of the lattice period change [ru

Thorium-Based Fuels Preliminary Lattice Cell Studies for Candu Reactors

International Nuclear Information System (INIS)

Margeanu, C.A.; Rizoiu, A.C.

2009-01-01

The choice of nuclear power as a major contributor to the future global energy needs must take into account acceptable risks of nuclear weapon proliferation, in addition to economic competitiveness, acceptable safety standards, and acceptable waste disposal options. Candu reactors offer a proven technology, safe and reliable reactor technology, with an interesting evolutionary potential for proliferation resistance, their versatility for various fuel cycles creating premises for a better utilization of global fuel resources. Candu reactors impressive degree of fuel cycle flexibility is a consequence of its channel design, excellent neutron economy, on-power refueling, and simple fuel bundle. These features facilitate the introduction and exploitation of various fuel cycles in Candu reactors in an evolutionary fashion. The main reasons for our interest in Thorium-based fuel cycles have been, globally, to extend the energy obtainable from natural Uranium and, locally, to provide a greater degree of energy self-reliance. Applying the once through Thorium (OTT) cycle in existing and advanced Candu reactors might be seen as an evaluative concept for the sustainable development both from the economic and waste management points of view. Two Candu fuel bundles project will be used for the proposed analysis, namely the Candu standard fuel bundle with 37 fuel elements and the CANFLEX fuel bundle with 43 fuel elements. Using the Canadian proposed scheme - loading mixed ThO 2 -SEU CANFLEX bundles in Candu 6 reactors - simulated at lattice cell level led to promising conclusions on operation at higher fuel burnups, reduction of the fissile content to the end of the cycle, minor actinide content reduction in the spent fuel, reduction of the spent fuel radiotoxicity, presence of radionuclides emitting strong gamma radiation for proliferation resistance benefit. The calculations were performed using the lattice codes WIMS and Dragon (together with the corresponding nuclear data

Synthesis and phase transformation mechanism of Nb{sub 2}C carbide phases

Energy Technology Data Exchange (ETDEWEB)

Vishwanadh, B., E-mail: visubathula@gmail.com [Materials Science Division, Bhabha Atomic Research Centre, Mumbai 400 094 (India); Murthy, T.S.R.Ch. [Materials Processing Division, Bhabha Atomic Research Centre, Mumbai 400 094 (India); Arya, A.; Tewari, R.; Dey, G.K. [Materials Science Division, Bhabha Atomic Research Centre, Mumbai 400 094 (India)

2016-06-25

In the present work, Niobium carbide samples were prepared through powder metallurgy route using spark plasma sintering technique. Some of these samples were heat treated at 900 °C up to 7 days. In order to investigate the phase transformation in Nb{sub 2}C carbide, the as-prepared and heat treated samples were characterized by X-ray diffraction, scanning electron microscopy and electron back scattered diffraction (EBSD) and transmission electron microscopy techniques. EBSD could index the same area of the sample in terms of any of the three allotropes of Nb{sub 2}C carbide phases (γ-Nb{sub 2}C, β-Nb{sub 2}C and α-Nb{sub 2}C) with good confidence index. From the EBSD patterns orientation relationships (OR) among γ, β and α-Nb{sub 2}C have been determined. Based on this OR when crystals of the three allotropes were superimposed, it has revealed that the basic Nb metal atom lattice (hcp lattice) in all the Nb{sub 2}C phases is same. The only difference exists in the carbides is the ordering of carbon atoms and vacancies in the octahedral positions of the hcp Nb metal atom lattice. Crystallographic analysis showed that for the transformation of γ-Nb{sub 2}C → β-Nb{sub 2}C → α-Nb{sub 2}C, large movement of Nb atoms is not required; but only by ordering of carbon atoms ensues the phase transformation. Literature shows that in the Nb–C system formation of the α-Nb{sub 2}C is not well established. Therefore, first principle calculations were carried out on these carbides. It revealed that the formation energy for α-Nb{sub 2}C is lower than the β and γ-Nb{sub 2}C carbides which indicate that the formation of α-Nb{sub 2}C is thermodynamically feasible. - Highlights: • Nb{sub 2}C carbide was produced by Spark Plasma Sintering in a single process. • Phase transformation mechanism of different Nb{sub 2}C carbide phases is studied. • In all the three Nb{sub 2}C carbides (γ, β, α), the base Nb lattice remains same. • Among γ, β and α-Nb{sub 2}C

Determination of space-energy distribution of resonance neutrons in reactor lattice cell and calculation of resonance integrals

International Nuclear Information System (INIS)

Zmijarevic, I.

1980-01-01

Space-energy distribution of resonance neutrons in reactor lattice cell was determined by solving the Boltzmann equation by spherical harmonics method applying P-3 approximation. Computer code SPLET used for these calculations is described. Resonance absorption and calculation of resonance integrals are described as well. Effective resonance integral values for U-238 resonance at 6.7 Ev are calculated for heavy water reactor cell with metal, oxide and carbide fuel elements

Evolutionary games combining two or three pair coordinations on a square lattice

Science.gov (United States)

Király, Balázs; Szabó, György

2017-10-01

We study multiagent logit-rule-driven evolutionary games on a square lattice whose pair interactions are composed of a maximal number of nonoverlapping elementary coordination games describing Ising-type interactions between just two of the available strategies. Using Monte Carlo simulations we investigate the macroscopic noise-level-dependent behavior of the two- and three-pair games and the critical properties of the continuous phase transtitions these systems exhibit. The four-strategy game is shown to be equivalent to a system that consists of two independent and identical Ising models.

Sensitivity analysis on various parameters for lattice analysis of DUPIC fuel with WIMS-AECL code

Energy Technology Data Exchange (ETDEWEB)

Roh, Gyu Hong; Choi, Hang Bok; Park, Jee Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

The code WIMS-AECL has been used for the lattice analysis of DUPIC fuel. The lattice parameters calculated by the code is sensitive to the choice of number of parameters, such as the number of tracking lines, number of condensed groups, mesh spacing in the moderator region, other parameters vital to the calculation of probabilities and burnup analysis. We have studied this sensitivity with respect to these parameters and recommend their proper values which are necessary for carrying out the lattice analysis of DUPIC fuel.

Sensitivity analysis on various parameters for lattice analysis of DUPIC fuel with WIMS-AECL code

Energy Technology Data Exchange (ETDEWEB)

Roh, Gyu Hong; Choi, Hang Bok; Park, Jee Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-12-31

The code WIMS-AECL has been used for the lattice analysis of DUPIC fuel. The lattice parameters calculated by the code is sensitive to the choice of number of parameters, such as the number of tracking lines, number of condensed groups, mesh spacing in the moderator region, other parameters vital to the calculation of probabilities and burnup analysis. We have studied this sensitivity with respect to these parameters and recommend their proper values which are necessary for carrying out the lattice analysis of DUPIC fuel.

Partially-reflected water-moderated square-piteched U(6.90)O₂ fuel rod lattices with 0.67 fuel to water volume ratio (0.800 CM Pitch)

Energy Technology Data Exchange (ETDEWEB)

Harms, Gary A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-09-01

The US Department of Energy (DOE) Nuclear Energy Research Initiative funded the design and construction of the Seven Percent Critical Experiment (7uPCX) at Sandia National Laboratories. The start-up of the experiment facility and the execution of the experiments described here were funded by the DOE Nuclear Criticality Safety Program. The 7uPCX is designed to investigate critical systems with fuel for light water reactors in the enrichment range above 5% ²³⁵U. The 7uPCX assembly is a water-moderated and -reflected array of aluminum-clad square-pitched U(6.90%)O₂ fuel rods.

Studies relating to construction materials to be used in different options for head end treatment in reprocessing of mixed carbide fuel of plutonium and uranium

International Nuclear Information System (INIS)

Rajan, S.K.; Palamalai, A.; Ravi, T.N.; Sampath, M.; Raman, V.R.; Balasubramanian, G.R.

1993-01-01

Mixed carbide of uranium and plutonium has been chosen as the fuel for the first core of Fast Breeder Test Reactor, installed in the Indira Gandhi Centre for Atomic Research. Reprocessing of this fuel is one of the vital steps to prove the viability of the fuel cycle. The head end treatment process introduces constraints in the reprocessing of carbide fuel when compared to the commonly used mixed oxide fuel. Three head end processes, namely direct oxidation, pyrohydrolysis and direct dissolution in nitric acid with oxidation of organic acids were considered for study for exercising the choice. The paper briefly describes the three processes. In each process probable material of construction and related problems are discussed. (author). 3 refs, 5 figs, 7 tabs

Maximal Fermi walk configurations on the directed square lattice and standard Young tableaux

International Nuclear Information System (INIS)

Arrowsmith, D K; Bhatti, F M; Essam, J W

2010-01-01

We consider configurations of n walkers each of which starts at the origin of a directed square lattice and makes the same number t of steps from node to node along the edges of the lattice. Bose walkers are not allowed to cross, but can share edges. Fermi walk configurations must satisfy the additional constraint that no two walkers traverse the same path. Since, for given t, there are only a finite number of t-step paths, there is a limit n max on the number of walkers allowed by the Fermi condition. The value of n max is determined for six types of boundary conditions. The number of Fermi configurations of n max walkers is also determined using a bijection to standard Young tableaux. In four cases there is no constraint on the endpoints of the walks and the relevant tableaux are shifted.

Preliminary Development of the MARS/FREK Spatial Kinetics Coupled System Code for Square Fueled Fast Reactor Applications

International Nuclear Information System (INIS)

Bae, Moo Hoon; Joo, Han Gyu

2009-01-01

Incorporation of a three-dimensional (3-D) reactor kinetics model into a system thermal-hydraulic (T/H) code enhances the capability to perform realistic analyses of the core neutronic behavior and the plant system dynamics which are coupled each other. For this advantage, several coupled system T/H and spatial kinetics codes, such as RELAP/PARCS, RELAP5/ PANBOX, and MARS/MASTER have been developed. These codes, however, so far limited to LWR applications. The objective of this work is to develop such a coupled code for fast reactor applications. Particularly, applications to lead-bismuth eutectic (LBE) cooled fast reactor are of interest which employ open square lattices. A fast reactor kinetics code applicable to square fueled cores called FREK is coupled the LBE version of the MARS code. The MARS/MASTER coupled code is used as the reference for the integration. The coupled code MARS/FREK is examined for a conceptual reactor called P-DEMO which is being developed by NUTRECK. In order to check the validity of the coupled code, however, the OECD MSLB benchmark exercise III calculation is solved first

Tunable all-angle negative refraction and photonic band gaps in two-dimensional plasma photonic crystals with square-like Archimedean lattices

International Nuclear Information System (INIS)

Zhang, Hai-Feng; Liu, Shao-Bin; Jiang, Yu-Chi

2014-01-01

In this paper, the tunable all-angle negative refraction and photonic band gaps (PBGs) in two types of two-dimensional (2D) plasma photonic crystals (PPCs) composed of homogeneous plasma and dielectric (GaAs) with square-like Archimedean lattices (ladybug and bathroom lattices) for TM wave are theoretically investigated based on a modified plane wave expansion method. The type-1 structure is dielectric rods immersed in the plasma background, and the complementary structure is named as type-2 PPCs. Theoretical simulations demonstrate that the both types of PPCs with square-like Archimedean lattices have some advantages in obtaining the higher cut-off frequency, the larger PBGs, more number of PBGs, and the relative bandwidths compared to the conventional square lattices as the filling factor or radius of inserted rods is same. The influences of plasma frequency and radius of inserted rod on the properties of PBGs for both types of PPCs also are discussed in detail. The calculated results show that PBGs can be manipulated by the parameters as mentioned above. The possibilities of all-angle negative refraction in such two types of PPCs at low bands also are discussed. Our calculations reveal that the all-angle negative phenomena can be observed in the first two TM bands, and the frequency range of all-angle negative refraction can be tuned by changing plasma frequency. Those properties can be used to design the optical switching and sensor

Modeling of MHD natural convection in a square enclosure having an adiabatic square shaped body using Lattice Boltzmann Method

Directory of Open Access Journals (Sweden)

Ahmed Kadhim Hussein

2016-03-01

Full Text Available A steady laminar two-dimensional magneto-hydrodynamics (MHD natural convection flow in a square enclosure filled with an electrically conducting fluid is numerically investigated using Lattice Boltzmann Method (LBM. The left and right vertical sidewalls of the square enclosure are maintained at hot and cold temperatures respectively. The horizontal top and bottom walls are considered thermally insulated. An adiabatic square shaped body is located in the center of a square enclosure and an external magnetic field is applied parallel to the horizontal x-axis. In the present work, the following parametric ranges of the non-dimensional groups are utilized: Hartmann number is varied as 0 ⩽ Ha ⩽ 50, Rayleigh number is varied as 103 ⩽ Ra ⩽ 105, Prandtl number is varied 0.05 ⩽ Pr ⩽ 5. It is found that the Hartmann number, Rayleigh number, and Prandtl number have an important role on the flow and thermal characteristics. It is found that when the Hartmann number increases the average Nusselt number decreases. The results also explain that the effect of magnetic field on flow field increases by increasing Prandtl number. However, the Prandtl number effect on the average Nusselt number with a magnetic field is less than the case without a magnetic field. Comparisons with previously published numerical works are performed and good agreements between the results are observed.

Gapped paramagnetic state in a frustrated spin-1/2 Heisenberg antiferromagnet on the cross-striped square lattice

Science.gov (United States)

Li, P. H. Y.; Bishop, R. F.

2018-03-01

We implement the coupled cluster method to very high orders of approximation to study the spin-1/2 J1 -J2 Heisenberg model on a cross-striped square lattice. Every nearest-neighbour pair of sites on the square lattice has an isotropic antiferromagnetic exchange bond of strength J1 > 0 , while the basic square plaquettes in alternate columns have either both or neither next-nearest-neighbour (diagonal) pairs of sites connected by an equivalent frustrating bond of strength J2 ≡ αJ1 > 0 . By studying the magnetic order parameter (i.e., the average local on-site magnetization) in the range 0 ≤ α ≤ 1 of the frustration parameter we find that the quasiclassical antiferromagnetic Néel and (so-called) double Néel states form the stable ground-state phases in the respective regions α α1bc = 0.615(5) . The double Néel state has Néel (⋯ ↑↓↑↓ ⋯) ordering along the (column) direction parallel to the stripes of squares with both or no J2 bonds, and spins alternating in a pairwise (⋯ ↑↑↓↓↑↑↓↓ ⋯) fashion along the perpendicular (row) direction, so that the parallel pairs occur on squares with both J2 bonds present. Further explicit calculations of both the triplet spin gap and the zero-field uniform transverse magnetic susceptibility provide compelling evidence that the ground-state phase over all or most of the intermediate regime α1ac < α < α1bc is a gapped state with no discernible long-range magnetic order.

Delayed Fission Product Gamma-Ray Transmission Through Low Enriched UO₂ Fuel Pin Lattices in Air

Energy Technology Data Exchange (ETDEWEB)

Trumbull, TH [Rensselaer Polytechnic Inst., Troy, NY (United States)

2004-10-18

The transmission of delayed fission-product gamma rays through various arrangements of low-enriched UO2 fuel pin lattices in an air medium was studied. Experimental measurements, point-kernel and Monte Carlo photon transport calculations were performed to demonstrate the shielding effect of ordered lattices of fuel pins on the resulting gamma-ray dose to a detector outside the lattice. The variation of the gamma-ray dose on the outside of the lattice as a function of radial position, the so-called “channeling” effect, was analyzed. Techniques for performing experimental measurements and data reduction at Rensselaer Polytechnic Institute’s Reactor Critical Facility (RCF) were derived. An experimental apparatus was constructed to hold the arrangements of fuel pins for the measurements. A gamma-ray spectroscopy system consisting of a sodium-iodide scintillation detector was used to collect data. Measurements were made with and without a collimator installed. A point-kernel transport code was developed to map the radial dependence of the gamma-ray flux. Input files for the Monte Carlo code, MCNP, were also developed to accurately model the experimental measurements. The results of the calculations were compared to the experimental measurements. In order to determine the delayed fission-product gamma-ray source for the calculations, a technique was developed using a previously written code, DELBG and the reactor state-point data obtained during the experimental measurements. Calculations were performed demonstrating the effects of material homogenization on the gamma-ray transmission through the fuel pin lattice.Homogeneous and heterogeneous calculations were performed for all RCF fuel pin lattices as well as for a typical commercial pressurized water reactor fuel bundle. The results of the study demonstrated the effectiveness of the experimental measurements to isolate the channeling effect of delayed fission-product gamma-rays through lattices of RCF fuel pins

Uncertainty Analysis of Light Water Reactor Fuel Lattices

Directory of Open Access Journals (Sweden)

C. Arenas

2013-01-01

Full Text Available The study explored the calculation of uncertainty based on available cross-section covariance data and computational tool on fuel lattice levels, which included pin cell and the fuel assembly models. Uncertainty variations due to temperatures changes and different fuel compositions are the main focus of this analysis. Selected assemblies and unit pin cells were analyzed according to the OECD LWR UAM benchmark specifications. Criticality and uncertainty analysis were performed using TSUNAMI-2D sequence in SCALE 6.1. It was found that uncertainties increase with increasing temperature, while kinf decreases. This increase in the uncertainty is due to the increase in sensitivity of the largest contributing reaction of uncertainty, namely, the neutron capture reaction 238U(n, γ due to the Doppler broadening. In addition, three types (UOX, MOX, and UOX-Gd2O3 of fuel material compositions were analyzed. A remarkable increase in uncertainty in kinf was observed for the case of MOX fuel. The increase in uncertainty of kinf in MOX fuel was nearly twice the corresponding value in UOX fuel. The neutron-nuclide reaction of 238U, mainly inelastic scattering (n, n′, contributed the most to the uncertainties in the MOX fuel, shifting the neutron spectrum to higher energy compared to the UOX fuel.

Medium temperature reaction between lanthanide and actinide carbides and hydrogen

International Nuclear Information System (INIS)

Dean, G.; Lorenzelli, R.; Pascard, R.

1964-01-01

Hydrogen is fixed reversibly by the lanthanide and actinide mono carbides in the range 25 - 400 C, as for pure corresponding metals. Hydrogen goes into the carbides lattice through carbon vacancies and the total fixed amount is approximately equal to two hydrogen atoms per initial vacancy. Final products c.n thus be considered as carbo-hydrides of general formula M(C 1-x , H 2x ). The primitive CFC, NaCl type, structure remains unchanged but expands strongly in the case of actinide carbides. With lanthanide carbides, hydrogenation induces a phase transformation with reappearance of the metal structure (HCP). Hydrogen decomposition pressures of all the studied carbo-hydrides are greater than those of the corresponding di-hydrides. (authors) [fr

A study on the formation of uranium carbide in an induction furnace

International Nuclear Information System (INIS)

Song, In Young; Lee, Yoon Sang; Kim, Eung Soo; Lee, Don Bae; Kim, Chang Kyu

2005-01-01

Uranium is a typical carbide-forming element. Three carbides, UC, U 2 C 3 and UC 2 , are formed in the uranium-carbon system. The most important of these as fuel is uranium monocarbide UC. It is well known that Uranium carbides can be obtained by three basic methods: 1) by reaction of uranium metal with carbon; 2) by reaction of uranium metal powder with gaseous hydrocarbons; 3) by reaction of uranium oxides with carbon. The use of uranium monocarbide, or materials based on it, has great prospects as fuel for nuclear reactors. It is quite possible that uranium dicarbide UC 2 may also acquire great importance as a fuel, particularly in dispersion fuel elements with graphite matrix. In the present study, uranium carbides are obtained by direct reaction of uranium metal with graphite in a high frequency induction furnace

Solution of semi-flexible self-avoiding trails on a Husimi lattice built with squares

Science.gov (United States)

Oliveira, Tiago J.; Dantas, Wellington G.; Prellberg, Thomas; Stilck, Jürgen F.

2018-02-01

We study a model of semi-flexible self-avoiding trails, where the lattice paths are constrained to visit each lattice edge at most once, with configurations weighted by the number of collisions, crossings and bends, on a Husimi lattice built with squares. We find a rich phase diagram with five phases: a non-polymerised phase (NP), low density (P1) and high density (P2) polymerised phases, and, for sufficiently large stiffness, two additional anisotropic (nematic) (AN1 and AN2) polymerised phases within the P1 phase. Moreover, the AN1 phase which shows a broken symmetry with a preferential direction, is separated from the P1 phase by the other nematic AN2 phase. Although this scenario is similar to what was found in our previous calculation on the Bethe lattice, where the AN-P1 transition was discontinuous and critical, the presence of the additional nematic phase between them introduces a qualitative difference. Other details of the phase diagram are that a line of tri-critical points may separate the P1-P2 transition surface into a continuous and a discontinuous portion, and that the same may happen at the NP-P1 transition surface, details of which depend on whether crossings are allowed or forbidden. A critical end-point line is also found in the phase diagram.

Calculation of vapour pressures over mixed carbide fuels

International Nuclear Information System (INIS)

Joseph, M.; Mathews, C.K.

1988-01-01

Vapour pressure over the uranium-plutonium mixed carbide (Usub(l-p) Pusub(p C) was calculated in the temperature range of 1300-9000 for various compositions (p=0.1 to 0.7). Effects of variation of the sesquicarbide content were also studied. The principle of corresponding states was applied to UC and mixed carbides to obtain the equation of state. (author)

DANCOFF-3, Dancoff Correction for Cylindrical Fuel Rod at H2O Gaps and for Fuel Clusters

International Nuclear Information System (INIS)

Sauer, A.

1989-01-01

1 - Nature of physical problem solved: Calculation of the Dancoff correction for cylindrical fuel rods in square and hexagonal infinite lattices, for fuel element rods near water gaps, and for fuel rod clusters. 2 - Method of solution: Evaluation by direct numerical integration over the moderator region. 3 - Restrictions on the complexity of the problem: For every rod arrangement at most 100 cases with different materials cross- sections

Numerical analysis of the reactivity for the dry lattices above the water level of the critical fuel cores

International Nuclear Information System (INIS)

Nauchi, Yasushi; Kameyama, Takanori

2003-01-01

Criticality analysis has been performed for dozens of tank type cores in which fuel lattices are loaded vertically and partially immersed in light water. The reactivity effect of dry part of lattices stuck above the critical water level has been calculated using the continuous energy Monte Carlo method. The reactivity effect exceeds 0.8% both for MOX and UOX fuel lattices of large buckling (B z 2 > 0.0025 cm -2 ). It is evaluated that at least 20 cm length of fuel rods above the critical water level has significant reactivity effect. (author)

Apparatus for surface treatment of U-Pu carbide fuel samples

International Nuclear Information System (INIS)

Fukushima, Susumu; Arai, Yasuo; Handa, Muneo; Ohmichi, Toshihiko; Shiozawa, Ken-ichi.

1979-05-01

Apparatus has been constructed for treating the surface of U-Pu carbide fuel samples for EPMA. The treatment is to clean off oxide layer on the surface, then coat with an electric-conductive material. The apparatus, safe in handling plutonium, operates as follows. (1) To avoid oxidation of the analyzing surface by oxygen and water in the air, series of cleaning and coating, i.e. ion-etching and ion-coating or ion-etching and vacuum-evaporation is done at the same time in an inert gas atmosphere. (2) Ion-etching is possible on samples embedded in non-electric-conductive and low heat-conductive resin. (3) Since the temperature rise in (2) is negligible, there is no deterioration of the samples. (author)

Bond percolation in a square lattice in presence of a 'magnetic field'

International Nuclear Information System (INIS)

Oliveira, P.M.C. de; Queiroz, S.L.A. de; Riera, R.; Chaves, C.M.G.F.

1979-10-01

A calculation of the bond percolation problem in a square lattice in presence of a magnetic field is presented using the position space renormalization group and cells of dimension b x b, where b runs from 2 up to 5. Due to symmetry, the calculation splits into two parts, one determining the 'thermal' exponent ν and the other, the magnetic exponent eta. For the largest cell in each case, we get ν = 1.355 (b=5) and eta = 0.244 (b=4), in good agreement with series results of Dunn et al. Comments are made on the extrapolation of the results to b = infinity. (Author) [pt

Process for the manufacture of a fuel catalyst made of tungsten carbide for electrochemical fuel cells. Verfahren zur Herstellung eines Brennstoffkatalysators aus Wolframcarbid fuer elektrochemische Brennstoffzellen

Energy Technology Data Exchange (ETDEWEB)

Baresel, D.; Gellert, W.; Scharner, P.

1982-05-19

The invention refers to a process for the manufacture of a fuel catalyst made of tungsten carbide for the direct generation of electrical energy by the oxidation of hydrogen, formaldehyde or formic acid in electrochemical fuel cells. Tungsten carbide is obtained by carburisation of tungsten or tungsten oxide by carbon monoxide. The steps of the process are as follows: dissolving the commercial-quality tungstic acid in ammonium hydroxide; precipitating the tungstic acid with concentrated hydrochloric acid; drying in a vacuum and then heating to 200/sup 0/C to remove the water of crystallisation forming tungsten trioxide; and mixing the tungsten trioxide with zinc powder and heating to 600/sup 0/C. The zinc oxide is dissolved with hydrochloric acid after cooling. The finely divided tungsten obtained in this way is converted with carbon monoxide in a quartz tube at 700/sup 0/C.

High resolution electron back-scatter diffraction analysis of thermally and mechanically induced strains near carbide inclusions in a superalloy

Energy Technology Data Exchange (ETDEWEB)

Karamched, Phani S., E-mail: phani.karamched@materials.ox.ac.uk [Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom); Wilkinson, Angus J. [Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom)

2011-01-15

Cross-correlation-based analysis of electron back-scatter diffraction (EBSD) patterns has been used to obtain high angular resolution maps of lattice rotations and elastic strains near carbides in a directionally solidified superalloy MAR-M-002. Lattice curvatures were determined from the EBSD measurements and used to estimate the distribution of geometrically necessary dislocations (GNDs) induced by the deformation. Significant strains were induced by thermal treatment due to the lower thermal expansion coefficient of the carbide inclusions compared to that of the matrix. In addition to elastic strains the mismatch was sufficient to have induced localized plastic deformation in the matrix leading to a GND density of 3 x 10{sup 13} m{sup -2} in regions around the carbide. Three-point bending was then used to impose strain levels within the range {+-}12% across the height of the bend bar. EBSD lattice curvature measurements were then made at both carbide-containing and carbide-free regions at different heights across the bar. The average GND density increases with the magnitude of the imposed strain (both in tension and compression), and is markedly higher near the carbides particles. The higher GND densities near the carbides (order of 10{sup 14} m{sup -2}) are generated by the large strain gradients produced around the plastically rigid inclusion during mechanical deformation with some minor contribution from the pre-existing residual deformation caused by the thermal mismatch between carbide and nickel matrix.

Non-locality and the flux line lattice square to hexagonal symmetry transition in the borocarbide superconductors

DEFF Research Database (Denmark)

Eskildsen, M.R.; Fisher, I.R.; Gammel, P.L.

2000-01-01

Using small angle neutron scattering we have studied the square to hexagonal flux line lattice symmetry transition in different members of the borocarbide superconductors. The studies were performed using samples of ErNi2B2C, Lu(Ni1-xCox)(2)B2C with cobalt doping levels x = 1.5-9% and Y0.64Lu0.36Ni...

Studies on the structure of zirconium carbide powders subjected to vibration grinding

International Nuclear Information System (INIS)

Kravchik, A.E.; Neshpor, V.S.

1976-01-01

The present work is a study of zirconium carbide powders subjected to vibratory milling in various media. The powders were comminuted in air (dry milling), benzene, trichloroethylene, and distilled water. The milling time was 10-160 h. The chemical compositions, specific surfaces, and crystal lattice parameters of the powder in the initial condition and after milling for 100 h in the various media are given. Vibratory milling of zirconium carbide powder can be successfully performed in benzene. Comminution in benzene enables a large specific surface to be attained, with practically no chemical reaction between the medium and the milling products. In milling in trichloroethylene the latter decomposes, with the formation of hydrochloric acid which reacts with the milling products. In a study of the fine structure parameters of zirconium carbide in the , , and directions the smallest crystal lattice strains and block sizes were observed in the direction. This may be taken as evidence that under such disintegration conditions the (110) planes constitute cleavage planes. An evaluation of internal and surface energies established that the strained crystal lattice energy reaches values which must be allowed for in any subsequent uses of the powder

A practical optimization procedure for radial BWR fuel lattice design using tabu search with a multiobjective function

International Nuclear Information System (INIS)

Francois, J.L.; Martin-del-Campo, C.; Francois, R.; Morales, L.B.

2003-01-01

An optimization procedure based on the tabu search (TS) method was developed for the design of radial enrichment and gadolinia distributions for boiling water reactor (BWR) fuel lattices. The procedure was coded in a computing system in which the optimization code uses the tabu search method to select potential solutions and the HELIOS code to evaluate them. The goal of the procedure is to search for an optimal fuel utilization, looking for a lattice with minimum average enrichment, with minimum deviation of reactivity targets and with a local power peaking factor (PPF) lower than a limit value. Time-dependent-depletion (TDD) effects were considered in the optimization process. The additive utility function method was used to convert the multiobjective optimization problem into a single objective problem. A strategy to reduce the computing time employed by the optimization was developed and is explained in this paper. An example is presented for a 10x10 fuel lattice with 10 different fuel compositions. The main contribution of this study is the development of a practical TDD optimization procedure for BWR fuel lattice design, using TS with a multiobjective function, and a strategy to economize computing time

Automatic fuel lattice design in a boiling water reactor using a particle swarm optimization algorithm and local search

International Nuclear Information System (INIS)

Lin Chaung; Lin, Tung-Hsien

2012-01-01

Highlights: ► The automatic procedure was developed to design the radial enrichment and gadolinia (Gd) distribution of fuel lattice. ► The method is based on a particle swarm optimization algorithm and local search. ► The design goal were to achieve the minimum local peaking factor. ► The number of fuel pins with Gd and Gd concentration are fixed to reduce search complexity. ► In this study, three axial sections are design and lattice performance is calculated using CASMO-4. - Abstract: The axial section of fuel assembly in a boiling water reactor (BWR) consists of five or six different distributions; this requires a radial lattice design. In this study, an automatic procedure based on a particle swarm optimization (PSO) algorithm and local search was developed to design the radial enrichment and gadolinia (Gd) distribution of the fuel lattice. The design goals were to achieve the minimum local peaking factor (LPF), and to come as close as possible to the specified target average enrichment and target infinite multiplication factor (k ∞ ), in which the number of fuel pins with Gd and Gd concentration are fixed. In this study, three axial sections are designed, and lattice performance is calculated using CASMO-4. Finally, the neutron cross section library of the designed lattice is established by CMSLINK; the core status during depletion, such as thermal limits, cold shutdown margin and cycle length, are then calculated using SIMULATE-3 in order to confirm that the lattice design satisfies the design requirements.

Fermionic Spinon Theory of Square Lattice Spin Liquids near the Néel State

Directory of Open Access Journals (Sweden)

Alex Thomson

2018-01-01

Full Text Available Quantum fluctuations of the Néel state of the square lattice antiferromagnet are usually described by a CP^{1} theory of bosonic spinons coupled to a U(1 gauge field, and with a global SU(2 spin rotation symmetry. Such a theory also has a confining phase with valence bond solid (VBS order, and upon including spin-singlet charge-2 Higgs fields, deconfined phases with Z_{2} topological order possibly intertwined with discrete broken global symmetries. We present dual theories of the same phases starting from a mean-field theory of fermionic spinons moving in π flux in each square lattice plaquette. Fluctuations about this π-flux state are described by (2+1-dimensional quantum chromodynamics (QCD_{3} with a SU(2 gauge group and N_{f}=2 flavors of massless Dirac fermions. It has recently been argued by Wang et al. [Deconfined Quantum Critical Points: Symmetries and Dualities, Phys. Rev. X 7, 031051 (2017.PRXHAE2160-330810.1103/PhysRevX.7.031051] that this QCD_{3} theory describes the Néel-VBS quantum phase transition. We introduce adjoint Higgs fields in QCD_{3} and obtain fermionic dual descriptions of the phases with Z_{2} topological order obtained earlier using the bosonic CP^{1} theory. We also present a fermionic spinon derivation of the monopole Berry phases in the U(1 gauge theory of the VBS state. The global phase diagram of these phases contains multicritical points, and our results imply new boson-fermion dualities between critical gauge theories of these points.

Fuel assemblies for use in BWR type reactors

International Nuclear Information System (INIS)

Hirukawa, Koji.

1987-01-01

Purpose: To moderate the peak configuration of the burnup degree change curve for the infinite multiplication factor by applying an improvement to the arrangement of fuel rods. Constitution: In a fuel assembly for a BWR type reactor comprising a plurality of fuel rods and water rods arranged in a square lattice, fuel rods containing burnable poisons are arranged at four corners at the second and the third layers from the outside of the square lattice arrangement. Among them, the Cd poison effect in the burnable poison incorporated fuel rods disposed at the second layer is somewhat greater at the initial burning stage and then rapidly decreased along with burning. While on the other hand, the poison effect of the burnable poison-incorporated fuel rods at the third layer is smaller than that at the second layer at the initial burning stage and the reduction in the poison effect due to burning is somewhat more moderate. Since these fuel rods are in adjacent with each other, they interfere to each other and also provide an effect of moderating the burning of the burnable poisons. (Takahashi, M.)

Self-avoiding walk on a square lattice with correlated vacancies

Science.gov (United States)

Cheraghalizadeh, J.; Najafi, M. N.; Mohammadzadeh, H.; Saber, A.

2018-04-01

The self-avoiding walk on the square site-diluted correlated percolation lattice is considered. The Ising model is employed to realize the spatial correlations of the metric space. As a well-accepted result, the (generalized) Flory's mean-field relation is tested to measure the effect of correlation. After exploring a perturbative Fokker-Planck-like equation, we apply an enriched Rosenbluth Monte Carlo method to study the problem. To be more precise, the winding angle analysis is also performed from which the diffusivity parameter of Schramm-Loewner evolution theory (κ ) is extracted. We find that at the critical Ising (host) system, the exponents are in agreement with Flory's approximation. For the off-critical Ising system, we find also a behavior for the fractal dimension of the walker trace in terms of the correlation length of the Ising system ξ (T ) , i.e., DFSAW(T ) -DFSAW(Tc) ˜1/√{ξ (T ) } .

Thermohydraulic analysis of BWR and PWR spent fuel assemblies contained within square canisters

International Nuclear Information System (INIS)

Wiles, L.E.; McCann, R.A.

1981-09-01

This report presents the results of several thermohydraulic simulations of spent fuel assembly/canister configurations performed in support of a program investigating the feasibility of storing spent nuclear fuel assemblies in canisters that would be stored in an air environment. Eleven thermohydraulic simulations were performed. Five simulations were performed using a single BWR fuel assembly/canister design. The various cases were defined by changing the canister spacing and the heat generation rate of the fuel assembly. For each simulation a steady-state thermohydraulic solution was achieved for the region inside the canister. Similarly, six simulations were performed for a single PWR fuel assembly/canister design. The square fuel rod arrays were contained in square canisters which would permit closer packing of the canisters in a storage facility. However, closer packing of the canisters would result in higher fuel temperatures which would possibly have an adverse impact on fuel integrity. Thus, the most important aspect of the analysis was to define the peak fuel assembly temperatures for each case. These results are presented along with various temperature profiles, heat flux distributions, and air velocity profiles within the canister. 48 figures, 4 tables

Computational Design of Advanced Nuclear Fuels

International Nuclear Information System (INIS)

Savrasov, Sergey; Kotliar, Gabriel; Haule, Kristjan

2014-01-01

The objective of the project was to develop a method for theoretical understanding of nuclear fuel materials whose physical and thermophysical properties can be predicted from first principles using a novel dynamical mean field method for electronic structure calculations. We concentrated our study on uranium, plutonium, their oxides, nitrides, carbides, as well as some rare earth materials whose 4f eletrons provide a simplified framework for understanding complex behavior of the f electrons. We addressed the issues connected to the electronic structure, lattice instabilities, phonon and magnon dynamics as well as thermal conductivity. This allowed us to evaluate characteristics of advanced nuclear fuel systems using computer based simulations and avoid costly experiments.

Measurements of thermal disadvantage factors in light-water moderated PuO2-UO2 and UO2 lattices

International Nuclear Information System (INIS)

Ohno, Akio; Kobayashi, Iwao; Tsuruta, Harumichi; Hashimoto, Masao; Suzaki, Takenori

1980-01-01

The disadvantage factor for thermal neutrons in light-water moderated PuO 2 -UO 2 and UO 2 square lattices were obtained from measurements of thermal neutron density distributions in a unit lattice cell, measured with Dy-Al wire detectors. The lattices consisted of 3.4 w/o PuO 2 .UO 2 and 2.6 w/o UO 2 fuel rods, and the water-to-fuel volume ratio within the unit cell was parametrically changed. The PuO 2 .UO 2 and UO 2 fuel rods were designed to realize equal fissile atomic number density. The disadvantage factors thus measured were 1.36 +- 0.07, 1.37 +- 0.08, 1.40 +- 0.06 and 1.38 +- 0.06 in the PuO 2 .UO 2 fuel lattices, and 1.30 +- 0.06, 1.31 +- 0.08, 1.30 +- 0.08 and 1.33 +- 0.06 in the UO 2 , for water-to-fuel volume ratios, of 1.76, 2.00, 2.38 and 2.95, respectively. This difference in disadvantage factor between PuO 2 .UO 2 and UO 2 fuel lattices corresponds to about 8%. Calculated results obtained by multigroup transport code LASER agreed well with the measured ones. (author)

Synthesis of carbide fuels from nano-structured precursors: impact on carbo-reduction and physico-chemical properties

International Nuclear Information System (INIS)

Saravia, Alvaro

2015-01-01

The classical way classically used for manufacturing carbide fuels consists of carbo-reducing at high temperature (1600 C) and under primary vacuum a mixture of AnO 2 and graphite powders. These conditions are disadvantageous for the synthesis of mixed (U,Pu)C carbides on account of plutonium volatilization. Therefore, one of the main aims of these studies is to decrease the carbo-reduction temperature. The experiments focused mainly on the lowering of the uranium oxide temperature. This result has been obtained with the use of uranium oxide and carbon nano-structured precursors. To achieve this goal colloidal suspensions of uranium oxide have been prepared and stabilized by cellulosic ethers. Cellulosic ethers are both stabiliser for uranium oxide nanoparticles and carbon source for carbo-reduction. It has been shown that these precursors are more efficient for carbo-reduction than the standard precursors: a reduction of 300 C of carbo-reduction temperature has been obtained. The impact of these precursors on carbo-reduction and on physico-chemical properties as well as the structural and microstructural characterizations of the obtained carbides have been carried out. (author) [fr

Effect of lattice deformation on temperature fields and heat transfer in the fuel elements of characteristic zones for a model of fast reactor fuel assembly

International Nuclear Information System (INIS)

Zhukov, A.V.; Matyukhin, N.M.; Sviridenko, E.Ya.

1980-01-01

Given are the experimental results for temperature fields in the model assembly in nonribbed simulators of the BN-600-type reactor fuel elements in the course of deformation of the lattice caused by shifting of the central and peripheral (lateral, angular) fuel elements by the value of the gap between the fuel elements (the limiting case when the fuel elements touch each other along the whole length). An assembly consisting of 37 electroheated pipes arranged in a triangular lattice with a relative step of S/d=1.185 is used as a model. The experiments were carried out on the sodium stand at constant energy release along the length of the fuel element simulators and at the Pe number changing in the 14-700 range. The data obtained show considerable increase of nonuniformities of the fuel element temperatures for characteristic zones of the fuel cassette assembly models of the fast reactor at deviations of the lattice geometric sizes from the nominal ones. For the central nonribbed element the temperature nonuniformity increases approximately 7.5 times and for the lateral element approximately 6 times when the elements touch each other along the whole length. The shift the central nonribbed element by the value of the gap between the fu.el elements leads to the decrease of heat transfer in comparison with heat transfer at the nominal geometry approximately 3-7 times in the 10-450 range for the Pe numbers. It is shown that the coolant temperature distribution along the assembly radius has a complex character (with a peak between the centre and the perifery) caused by redistribution of coolant consumptions due to fuel element lattice deformation

On change of vanadium carbide state during 20Kh3MVF steel heat treatment

International Nuclear Information System (INIS)

Gitgarts, M.I.; Maksimenko, V.N.

1975-01-01

The Xray diffraction study of vanadium carbide MC has been made in the steel-20KH3MVF quenched from 970 and 1040 deg and tempered at 660 deg for 210 hrs. It has been found that the constant of the MC crystal lattice regularly varies with the temperature of isothermal hold-up. In the steel tempered after quenching two vanadium carbides of different content could co-exist simultaneously: carbide formed in the quenching process and carbide formed during tempering. The discovered effect of the temperature dependence of the MC content is, evidently, inherent also to other steels containing vanadium

Symmetry breaking states of Bose-Einstein condensates in 1D double square well and optical lattice well

International Nuclear Information System (INIS)

Yuan Qingxin; Ding Guohui

2005-01-01

We investigate the phenomena of symmetry breaking and phase transition in the ground state of Bose-Einstein condensates (BECs). For BECs trapped in a double square well potential, we present symmetric and asymmetric ground states by using standing-wave expansion method. For BECs trapped in an optical lattice well potential (created by a standing laser wave, and not just an extension of the double square well potential), we reveal a phase transition by using plane-wave expansion method. At the same time we also study the ground state properties with changing the depth of potential and atomic interactions (restrict ourselves to the attractive regime)

Positron annihilation spectroscopy study of lattice defects in non-irradiated doped and un-doped fuels

Directory of Open Access Journals (Sweden)

Chollet Mélanie

2017-01-01

Full Text Available Fission gas behavior within the fuel structure plays a major role for the safety of nuclear fuels during operation in the nuclear power plant. Fission gas distribution and retention is determined by both, micro- and lattice-structure of the fuel matrix. The ADOPT (Advanced Doped Pellet Technology fuel, containing chromium and aluminum additives, shows larger grain sizes than standard (undoped UO2 fuel, enhancing the fission gas retention properties of the matrix. However, the additions of such trivalent cations shall also induce defects in the lattice. In this study, we investigated the microstructure of such doped fuels as well as a reference standard UO2 by positron annihilation spectroscopy (PAS. Although this technique is particularly sensitive to lattice point defects in materials, a wider application in the UO2 research is still missing. The PAS-lifetime components were measured in the hotlab facility of PSI using a 22Na source sandwiched between two 500-μm-thin sample discs. The values of lifetime at the center and the rim of both samples, examined to check at the radial homogeneity of the pellets, are not significantly different. The mean lifetimes were found to be longer in the ADOPT material, 220 ps, than in standard UO2, 190 ps, which indicates a larger presence of additional defects, presumably generated by the dopants. While two-component decomposition (bulk + one defect component could be performed for the standard material, only one lifetime component was found in the doped material. The absence of the bulk component in the ADOPT sample refers to a saturated positron trapping (i.e., all positrons are trapped at defects. In order to associate a type of lattice defect to each PAS component, interpretation of the PAS experimental observations was conducted with respect to existing experimental and modeling studies. This work has shown the efficiency of PAS to detect lattice point defects in UO2 produced by Cr and Al oxides

Galilean invariant lattice Boltzmann scheme for natural convection on square and rectangular lattices

NARCIS (Netherlands)

Sman, van der R.G.M.

2006-01-01

In this paper we present lattice Boltzmann (LB) schemes for convection diffusion coupled to fluid flow on two-dimensional rectangular lattices. Via inverse Chapman-Enskog analysis of LB schemes including source terms, we show that for consistency with physics it is required that the moments of the

Candidate for a fully frustrated square lattice in a verdazyl-based salt

Science.gov (United States)

Yamaguchi, H.; Tamekuni, Y.; Iwasaki, Y.; Hosokoshi, Y.

2018-05-01

We present an experimental realization of an S =1 /2 fully frustrated square lattice (FFSL) composed of a verdazyl-based salt (p -MePy-V) (TCNQ ) .(CH3)2CO . Ab initio molecular orbital calculations indicate that there are four types of competing ferro- and antiferromagnetic nearest-neighbor interactions present in the system, which combine to form an S =1 /2 FFSL. Below room temperature, the magnetic susceptibility of the material can be considered to arise from the S =1 /2 FFSL formed by the p -MePy-V and indicates that the system forms a quantum valence-bond solid state whose excitation energy is gapped. Furthermore, we also observe semiconducting behavior arising from the one-dimensional chain structure of the TCNQ molecules.

Proton Exchange Membrane Fuel Cell Modelling Using Moving Least Squares Technique

Directory of Open Access Journals (Sweden)

Radu Tirnovan

2009-07-01

Full Text Available Proton exchange membrane fuel cell, with low polluting emissions, is a great alternative to replace the traditional electrical power sources for automotive applications or for small stationary consumers. This paper presents a numerical method, for the fuel cell modelling, based on moving least squares (MLS. Experimental data have been used for developing an approximated model of the PEMFC function of the current density, air inlet pressure and operating temperature of the fuel cell. The method can be applied for modelling others fuel cell sub-systems, such as the compressor. The method can be used for off-line or on-line identification of the PEMFC stack.

Effect of Ti additive on (Cr, Fe)7C3 carbide in arc surfacing layer and its refined mechanism

International Nuclear Information System (INIS)

Zhou Yefei; Yang Yulin; Yang Jian; Hao Feifei; Li Da; Ren Xuejun; Yang Qingxiang

2012-01-01

Arc surfacing layer of hypoeutectic high chromium cast iron (HCCI) expects refiner carbides in the microstructure to improve its mechanical properties. In this paper, Ti additive as a strong carbide forming element was added in the hypoeutectic HCCI arc surfacing layer. Microstructure of titaniferous hypoeutectic HCCI was studied by optical microscopy, X-ray diffraction and field emission scanning electronic microscopy with energy dispersive spectrometer. Furthermore, the M(M = Cr, Fe) 7 C 3 carbide refinement mechanism was explained by the phase diagram calculation and lattice misfit theory. The results show that, the M 7 C 3 carbide in arc surfacing microstructure of hypoeutectic HCCI has been refined with 2 wt.% Ti additive, and TiC carbide can be observed in/around the M 7 C 3 carbide. With Ti addictive increasing, the micro-hardness along the depth in profile section of layer becomes more uniform, and the wear resistance has been improved. According to the phase diagram calculation, MC carbide precipitates prior to M 7 C 3 carbide in Fe-Cr-C-Ti alloy. In addition, the lattice misfit between (1 1 0) TiC and (010) Cr 7 C 3 is 9.257%, which indicates that the TiC as heterogeneous nuclei of the M 7 C 3 is medium effective. Therefore, the M 7 C 3 carbide can be refined.

Fluoride-Salt-Cooled High-Temperature Reactor (FHR) with Silicon-Carbide-Matrix Coated-Particle Fuel

International Nuclear Information System (INIS)

Forsberg, C. W.; Snead, Lance Lewis; Katoh, Yutai

2012-01-01

The FHR is a new reactor concept that uses coated-particle fuel and a low-pressure liquid-salt coolant. Its neutronics are similar to a high-temperature gas-cooled reactor (HTGR). The power density is 5 to 10 times higher because of the superior cooling properties of liquids versus gases. The leading candidate coolant salt is a mixture of 7 LiF and BeF 2 (FLiBe) possessing a boiling point above 1300 C and the figure of merit ρC p (volumetric heat capacity) for the salt slightly superior to water. Studies are underway to define a near-term base-line concept while understanding longer-term options. Near-term options use graphite-matrix coated-particle fuel where the graphite is both a structural component and the primary neutron moderator. It is the same basic fuel used in HTGRs. The fuel can take several geometric forms with a pebble bed being the leading contender. Recent work on silicon-carbide-matrix (SiCm) coated-particle fuel may create a second longer-term fuel option. SiCm coated-particle fuels are currently being investigated for use in light-water reactors. The replacement of the graphite matrix with a SiCm creates a new family of fuels. The first motivation behind the effort is to take advantage of the superior radiation resistance of SiC compared to graphite in order to provide a stable matrix for hosting coated fuel particles. The second motivation is a much more rugged fuel under accident, repository, and other conditions.

Analysis of crystallite size and microdeformation crystal lattice the tungsten carbide milling in mill high energy; Analise do tamanho do cristalito e microdeformacao da rede cristalina do carbeto de tugstenio moidos em moinho de alta energia

Energy Technology Data Exchange (ETDEWEB)

Silva, F.T. da; Nunes, M.A.M. [Universidade Federal do Rio Grande do Norte (PPGCEM/UFRN), Natal (Brazil). Programa de Pos-Graduacao em Ciencia e Engenharia de Materiais; Oliveira, R.M.V. de; Silva, G.G. da [Instituto Federal do Rio Grande do Norte (IFRN), Natal (Brazil); Souza, C.P. de; Gomes, U.U. [Universidade Federal do Rio Grande do Norte (UFRN), Natal (Brazil)

2010-07-01

The tungsten carbide (WC) has wide application due to its properties like high melting point, high hardness, wear resistance, oxidation resistance and good electrical conductivity. The microstructural characteristics of the starting powders influences the final properties of the carbide. In this context, the use of nanoparticle powders is an efficient way to improve the final properties of the WC. The high energy milling stands out from other processes to obtain nanometric powders due to constant microstructural changes caused by this process. Therefore, the objective is to undertake an analysis of microstructural characteristics on the crystallite size and microdeformations of the crystal lattice using the technique of X-ray diffraction (XRD) using the Rietveld refinement. The results show an efficiency of the milling process to reduce the crystallite size, leading to a significant deformation in the crystal lattice of WC from 5h milling. (author)

Fracture and Residual Characterization of Tungsten Carbide Cobalt Coatings on High Strength Steel

National Research Council Canada - National Science Library

Parker, Donald S

2003-01-01

Tungsten carbide cobalt coatings applied via high velocity oxygen fuel thermal spray deposition are essentially anisotropic composite structures with aggregates of tungsten carbide particles bonded...

Simulations and measurements of adiabatic annular flows in triangular, tight lattice nuclear fuel bundle model

Energy Technology Data Exchange (ETDEWEB)

Saxena, Abhishek, E-mail: asaxena@lke.mavt.ethz.ch [ETH Zurich, Laboratory for Nuclear Energy Systems, Department of Mechanical and Process Engineering, Sonneggstrasse 3, 8092 Zürich (Switzerland); Zboray, Robert [Laboratory for Thermal-hydraulics, Nuclear Energy and Safety Department, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Prasser, Horst-Michael [ETH Zurich, Laboratory for Nuclear Energy Systems, Department of Mechanical and Process Engineering, Sonneggstrasse 3, 8092 Zürich (Switzerland); Laboratory for Thermal-hydraulics, Nuclear Energy and Safety Department, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland)

2016-04-01

High conversion light water reactors (HCLWR) having triangular, tight-lattice fuels bundles could enable improved fuel utilization compared to present day LWRs. However, the efficient cooling of a tight lattice bundle has to be still proven. Major concern is the avoidance of high-quality boiling crisis (film dry-out) by the use of efficient functional spacers. For this reason, we have carried out experiments on adiabatic, air-water annular two-phase flows in a tight-lattice, triangular fuel bundle model using generic spacers. A high-spatial-resolution, non-intrusive measurement technology, cold neutron tomography, has been utilized to resolve the distribution of the liquid film thickness on the virtual fuel pin surfaces. Unsteady CFD simulations have also been performed to replicate and compare with the experiments using the commercial code STAR-CCM+. Large eddies have been resolved on the grid level to capture the dominant unsteady flow features expected to drive the liquid film thickness distribution downstream of a spacer while the subgrid scales have been modeled using the Wall Adapting Local Eddy (WALE) subgrid model. A Volume of Fluid (VOF) method, which directly tracks the interface and does away with closure relationship models for interfacial exchange terms, has also been employed. The present paper shows first comparison of the measurement with the simulation results.

Numerical prediction of turbulent heat transfer augmentation in an annular fuel channel with two-dimensional square ribs

International Nuclear Information System (INIS)

Takase, Kazuyuki

1996-01-01

The square-ribbed fuel rod for high temperature gas-cooled reactors was developed in order to enhance the turbulent heat transfer in comparison with the standard fuel rod. To evaluate the heat transfer performance of the square-ribbed fuel rod, the turbulent heat transfer coefficients in an annular fuel channel with repeated two-dimensional square ribs were analyzed numerically on a fully developed incompressible flow using the k - ε turbulence model and the two-dimensional axisymmetrical coordinate system. Numerical analyses were carried out for a range of Reynolds numbers from 3000 to 20000 and ratios of square-rib pitch to height of 10, 20 and 40, respectively. The predicted values of the heat transfer coefficients agreed within an error of 10% for the square-rib pitch to height ratio of 10, 20% for 20 and 25% for 40, respectively, with the heat transfer empirical correlations obtained from the experimental data. It was concluded by the present study that the effect of the heat transfer augmentation by square ribs could be predicted sufficiently by the present numerical simulations and also a part of its mechanism could be explained by means of the change in the turbulence kinematic energy distribution along the flow direction. (author)

Investigation of fuel lattice pitch changes influence on reactor performance through evaluate the neutronic parameters

International Nuclear Information System (INIS)

Zareian Ronizi, F.; Fadaei, A.H.; Setayeshi, S.; Shahidi, A.R.

2015-01-01

Highlights: • One of the most complex issues that Nu-engineers deal with is the design of NR core. • Numerous factors in nuclear core design depend on Fuel-to-Moderator volume ratio. • Aim of this research is to investigate RX performance for different lattice pitches. - Abstract: Nuclear reactor core design is one of the most complex issues that nuclear engineers deal with. The number and complexity of effective parameters and their impact on reactor design, which makes the problem difficult to solve, require precise knowledge of these parameters and their influence on the reactor operation. Numerous factors in a nuclear reactor core design depend on the Fuel-to-Moderator volume ratio, V F /V M , in a fuel cell. This ratio can be modified by changing the lattice pitch which is the thickness of water channels between fuels plates while keeping fuel slab dimensions fixed. Cooling and moderating properties of water are affected by such a change in a reactor core, and hence some parameters related to these properties might be changed. The aim of this research is to provide the suitable knowledge for nuclear core designing. To reach this goal, the first operating core of Tehran Research Reactor (TRR) with different lattice pitches is simulated, and the effect of different lattice pitches on some parameters such as effective multiplication factor (K eff ), reactor life time, distribution of neutron flux and power density in the core, as well as moderator temperature and density coefficient of reactivity are evaluated. The nuclear reactor analysis code, MTR-PC package is employed to carry out the considered calculation. Finally, the results are presented in some tables and graphs that provide useful information for nuclear engineers in the nuclear reactor core design

Reactor irradiation effect on the physical-mechanical properties of zirconium carbides and niobium carbides

International Nuclear Information System (INIS)

Andrievskij, R.A.; Vlasov, K.P.; Shevchenko, A.S.; Lanin, A.G.; Pritchin, S.A.; Klyushin, V.V.; Kurushin, S.P.; Maskaev, A.S.

1978-01-01

A study has been made of the effect of the reactor radiation by a flux of neutrons 1.5x10 20 n/cm 2 (E>=1 meV) at radiation temperatures of 150 and 1100 deg C on the physico-mechanical properties of carbides of zirconium and niobium and their equimolar hard solution. A difference has been discovered in the behaviour of the indicated carbides under the effect of radiation. Under the investigated conditions of radiation the density of zirconium carbide is being decreased, while in the niobium carbide no actual volumetric changes occur. The increase of the lattice period in ZrC is more significant than in NbC. The electric resistance of ZrC is also changed more significantly than in the case of NbC, while for the microhardness a reverse relationship is observed. Strength and elasticity modulus change insignificantly in both cases. Resistance to crack formation shows a higher reduction for ZrC than for NbC, while the thermal strength shows an approximately similar increase. The equimolar hard solution of ZrC and NbC behaves to great extent similar to ZrC, although the change in electric resistance reminds of NbC while thermal strength changes differently. The study of the microstructure of the specimens has shown that radiation causes a large number of etching patterns-dislocations in NbC which are almost absent in ZrC

Reactivity and reaction rate measurements in U--D2O lattices with coaxial fuel

International Nuclear Information System (INIS)

Pellarin, D.J.; Morris, B.M.

1976-12-01

Integral reaction rate parameters, intracell thermal neutron flux profiles, and material bucklings were measured for D 2 O-moderated uniform lattices in the exponential facility at the Savannah River Laboratory. Two different slightly enriched coaxial uranium fuel assemblies were examined over a wide range of triangular lattice pitches. Integral parameters are reported for inner and outer fuel separately providing data for a more detailed and rigorous comparison with computation than has been previously available. Results are compared with RAHAB calculations using ENDF/B-IV cross sections. Large discrepancies in agreement between calculation and experiment, outside of experimental errors and uncertainties in the input cross sections, probably result from the resonance capture models used by RAHAB

Optimization of uranium carbide fabrication by carbothermic reduction with limited oxygen content

International Nuclear Information System (INIS)

Raveu, Gaelle

2014-01-01

Mixed carbides (U, Pu)C, are good fuel candidate for generation IV reactors because of their high fissile atoms density and excellent thermal properties for economical (more compact and efficient cores) and safety reasons (high melting margin). UC can be imagine as a surrogate material ror R and D studies on (U,Pu)C fuel behavior, because of their similar structures. The carbothermic reaction was used because it is the most studied and now consider for industrial process. However, it involves powders manipulation: in air, carbide can strongly react at room temperature and under controlled atmosphere it can absorb impurities. An inerted installation under Ar, BaGCARA, was therefore used. Process improvements were carried out, including the sintering atmosphere in order to evaluate the impact on the sample purity (about oxygen content). The original method by ion beam analysis was used to determine the surface composition (oxygen in-depth profiles in the first microns and stoichiometry). This oxygen analysis was set for the first time in carbonaceous materials. XRD analysis showed the formation of an intermediate compound during the carbothermic reaction and a better crystallization of the samples fabricated in BaGCARA. They also have a better microstructure, density, and visual appearance if compared to former samples. Vacuum sintering leads to a denser UC with fewer second phases if compared to Ar, Ar/H 2 or controlled PC atmospheres. However, it was not possible to analyze carbides without air contact which may impact their lattice parameter and lead to their deterioration. When the carbide is initially free of oxygen, it oxidizes faster, more intensely and heterogeneously. The mechanical stress induced between the grains lead to fracturing the material, to corrosion cracking and then a de-bonding of the material. A study of oxidation mechanisms would be interesting to validate and understand the evolution of the material in contact with oxygen. A study of the

Numerical prediction of augmented turbulent heat transfer in an annular fuel channel with repeated two-dimensional square ribs

International Nuclear Information System (INIS)

Takase, K.

1996-01-01

The square-ribbed fuel rod for high temperature gas-cooled reactors was designed and developed so as to enhance the turbulent heat transfer in comparison with the previous standard fuel rod. The turbulent heat transfer characteristics in an annular fuel channel with repeated two-dimensional square ribs were analysed numerically on a fully developed incompressible flow using the k-ε turbulence model and the two-dimensional axisymmetrical coordinate system. Numerical analyses were carried out under the conditions of Reynolds numbers from 3000 to 20000 and ratios of square-rib pitch to height of 10, 20 and 40 respectively. The predictions of the heat transfer coefficients agreed well within an error of 10% for the square-rib pitch to height ratio of 10, 20% for 20 and 25% for 40 respectively, with the heat transfer empirical correlations obtained from the experimental data due to the simulated square-ribbed fuel rods. Therefore it was found that the effect of heat transfer augmentation due to the square ribs could be predicted by the present numerical simulations and the mechanism could be explained by the change in the turbulence kinematic energy distribution along the flow direction. (orig.)

Gas chromatographic determination of Di-n-butyl phosphate in radioactive lean organic solvent of FBTR carbide fuel reprocessing

International Nuclear Information System (INIS)

Velavendan, P.; Ganesh, S.; Pandey, N.K.; Kamachi Mudali, U.; Natarajan, R.

2011-01-01

In the present work Di-n- butyl phosphate (DBP) a degraded product of Tri-n-butyl phosphate (TBP) formed by acid hydrolysis and radiolysis in the PUREX process was analyzed. Lean organic streams of different fuel burn-up FBTR carbide fuel reprocessing solution was determined by standard Gas Chromatographic technique. The method involves the conversion of non-volatile Di-n-butyl phosphate into volatile and stable derivatives by the action of diazomethane and then determined by Gas Chromatograph (GC). A calibration graph was made for DBP concentration range of 200-2000 ppm with correlation coefficient of 0.99587 and RSD 1.2 %. (author)

Advanced Characterization Techniques for Silicon Carbide and Pyrocarbon Coatings on Fuel Particles for High Temperature Reactors (HTR)

Energy Technology Data Exchange (ETDEWEB)

Basini, V.; Charollais, F. [CEA Cadarache, DEN/DEC/SPUA, BP 1, 13108 St Paul Lez Durance (France); Dugne, O. [CEA Marcoule, DEN/DTEC/SCGS BP 17171 30207 Bagnols sur Ceze (France); Garcia, C. [Laboratoire des Composites Thermostructuraux (LCTS), UMR CNRS 5801, 3 allee de La Boetie, 33600 Pessac (France); Perez, M. [CEA Grenoble DRT/DTH/LTH, 17 rue des Martyrs, 38054 Grenoble cedex 9 (France)

2008-07-01

Cea and AREVA NP have engaged an extensive research and development program on HTR (high temperature reactor) fuel. The improving of safety of (very) high temperature reactors (V/HTR) is based on the quality of the fuel particles. This requires a good knowledge of the properties of the four-layers TRISO particles designed to retain the uranium and fission products during irradiation or accident conditions. The aim of this work is to characterize exhaustively the structure and the thermomechanical properties of each unirradiated layer (silicon carbide and pyrocarbon coatings) by electron microscopy (SEM, TEM), selected area electronic diffraction (SEAD), thermo reflectance microscopy and nano-indentation. The long term objective of this study is to define pertinent parameters for fuel performance codes used to better understand the thermomechanical behaviour of the coated particles. (authors)

Steady- and transient-state analysis of fully ceramic microencapsulated fuel with randomly dispersed tristructural isotropic particles via two-temperature homogenized model-II: Applications by coupling with COREDAX

International Nuclear Information System (INIS)

Lee, Yoon Hee; Cho, Bum Hee; Cho, Nam Zin

2016-01-01

In Part I of this paper, the two-temperature homogenized model for the fully ceramic microencapsulated fuel, in which tristructural isotropic particles are randomly dispersed in a fine lattice stochastic structure, was discussed. In this model, the fuel-kernel and silicon carbide matrix temperatures are distinguished. Moreover, the obtained temperature profiles are more realistic than those obtained using other models. Using the temperature-dependent thermal conductivities of uranium nitride and the silicon carbide matrix, temperature-dependent homogenized parameters were obtained. In Part II of the paper, coupled with the COREDAX code, a reactor core loaded by fully ceramic microencapsulated fuel in which tristructural isotropic particles are randomly dispersed in the fine lattice stochastic structure is analyzed via a two-temperature homogenized model at steady and transient states. The results are compared with those from harmonic- and volumetric-average thermal conductivity models; i.e., we compare keff eigenvalues, power distributions, and temperature profiles in the hottest single channel at a steady state. At transient states, we compare total power, average energy deposition, and maximum temperatures in the hottest single channel obtained by the different thermal analysis models. The different thermal analysis models and the availability of fuel-kernel temperatures in the two-temperature homogenized model for Doppler temperature feedback lead to significant differences

Study of the lattice parameter evolution of PWR irradiated MOX fuel by X-Ray diffraction

International Nuclear Information System (INIS)

Clavier, B.

1995-01-01

Fuel irradiation leads to a swelling resulting from the formation of gaseous (Kr, Xe) or solid fission products which are found either in solution or as solid inclusions in the matrix. This phenomena has to be evaluated to be taken into account in fuel cladding Interaction. Fuel swelling was studied as a function of burn up by measuring the corresponding cell constant evolution by X-Ray diffraction. This study was realized on Mixed Oxide Fuels (MOX) irradiated in a Pressurized Water Reactor (PWR) at different burn-up for 3 initial Pu contents. Lattice parameter evolutions were followed as a function of burn-up for the irradiated fuel with and without an annealing thermal treatment. These experimental evolutions are compared to the theoretical evolutions calculated from the hard sphere model, using the fission product concentrations determined by the APPOLO computer code. Contribution of varying parameters influencing the unit cell value is discussed. Thermal treatment effects were checked by metallography, X-Ray diffraction and microprobe analysis. After thermal treatment, no structural change was observed but a decrease of the lattice parameter was measured. This modification results essentially from self-irradiation defect annealing and not from stoichiometry variations. Microprobe analysis showed that about 15% of the formed Molybdenum is in solid solution In the oxide matrix. Micrographs showed the existence of Pu packs in the oxide matrix which induces a broadening of diffraction lines. The RIETVELD method used to analyze the X-Ray patterns did not allow to characterize independently the Pu packs and the oxide matrix lattice parameters. Nevertheless, with this method, the presence of micro-strains in the irradiated nuclear fuel could be confirmed. (author)

Tribological performance of polycrystalline tantalum-carbide-incorporated diamond films on silicon substrates

Science.gov (United States)

Ullah, Mahtab; Rana, Anwar Manzoor; Ahmed, E.; Malik, Abdul Sattar; Shah, Z. A.; Ahmad, Naseeb; Mehtab, Ujala; Raza, Rizwan

2018-05-01

Polycrystalline tantalum-carbide-incorporated diamond coatings have been made on unpolished side of Si (100) wafer by hot filament chemical vapor deposition process. Morphology of the coatings has been found to vary from (111) triangular-facetted to predominantly (111) square-faceted by increasing the concentration of tantalum carbide. The results have been compared to those of a diamond reference coating with no tantalum content. An increase in roughness has been observed with the increase of tantalum carbide (TaC) due to change in morphology of the diamond films. It is noticed that roughness of the coatings increases as grains become more square-faceted. It is found that diamond coatings involving tantalum carbide are not as resistant as diamond films with no TaC content and the coefficient of friction for such coatings with microcrystalline grains can be manipulated to 0·33 under high vacuum of 10-7 Torr. Such a low friction coefficient value enhances tribological behavior of unpolished Si substrates and can possibly be used in sliding applications.

Point defects and transport properties in carbides

International Nuclear Information System (INIS)

Matzke, Hj.

1984-01-01

Carbides of transition metals and of actinides are interesting and technologically important. The transition-metal carbides (or carbonitrides) are extensively being used as hard materials and some of them are of great interest because of the high transition temperature for superconductivity, e.g. 17 K for Nb(C,N). Actinide carbides and carbonitrides, (U,Pu)C and (U,Pu)(C,N) are being considered as promising advanced fuels for liquid metal cooled fast breeder nuclear reactors. Basic interest exists in all these materials because of their high melting points (e.g. 4250 K for TaC) and the unusually broad range of homogeneity of nonstoichiometric compositions (e.g. from UCsub(0.9) to UCsub(1.9) at 2500 K). Interaction of point defects to clusters and short-range ordering have recently been studied with elastic neutron diffraction and diffuse scattering techniques, and calculations of energies of formation and interaction of point defects became available for selected carbides. Diffusion measurements also exist for a number of carbides, in particular for the actinide carbides. The existing knowledge is discussed and summarized with emphasis on informative examples of particular technological relevance. (Auth.)

Understanding the Irradiation Behavior of Zirconium Carbide

International Nuclear Information System (INIS)

Motta, Arthur; Sridharan, Kumar; Morgan, Dane; Szlufarska, Izabela

2013-01-01

Zirconium carbide (ZrC) is being considered for utilization in high-temperature gas-cooled reactor fuels in deep-burn TRISO fuel. Zirconium carbide possesses a cubic B1-type crystal structure with a high melting point, exceptional hardness, and good thermal and electrical conductivities. The use of ZrC as part of the TRISO fuel requires a thorough understanding of its irradiation response. However, the radiation effects on ZrC are still poorly understood. The majority of the existing research is focused on the radiation damage phenomena at higher temperatures (>450ee)C) where many fundamental aspects of defect production and kinetics cannot be easily distinguished. Little is known about basic defect formation, clustering, and evolution of ZrC under irradiation, although some atomistic simulation and phenomenological studies have been performed. Such detailed information is needed to construct a model describing the microstructural evolution in fast-neutron irradiated materials that will be of great technological importance for the development of ZrC-based fuel. The goal of the proposed project is to gain fundamental understanding of the radiation-induced defect formation in zirconium carbide and irradiation response by using a combination of state-of-the-art experimental methods and atomistic modeling. This project will combine (1) in situ ion irradiation at a specialized facility at a national laboratory, (2) controlled temperature proton irradiation on bulk samples, and (3) atomistic modeling to gain a fundamental understanding of defect formation in ZrC. The proposed project will cover the irradiation temperatures from cryogenic temperature to as high as 800ee)C, and dose ranges from 0.1 to 100 dpa. The examination of this wide range of temperatures and doses allows us to obtain an experimental data set that can be effectively used to exercise and benchmark the computer calculations of defect properties. Combining the examination of radiation

MC Carbide Characterization in High Refractory Content Powder-Processed Ni-Based Superalloys

Science.gov (United States)

Antonov, Stoichko; Chen, Wei; Huo, Jiajie; Feng, Qiang; Isheim, Dieter; Seidman, David N.; Sun, Eugene; Tin, Sammy

2018-04-01

Carbide precipitates in Ni-based superalloys are considered to be desirable phases that can contribute to improving high-temperature properties as well as aid in microstructural refinement of the material; however, they can also serve as crack initiation sites during fatigue. To date, most of the knowledge pertaining to carbide formation has originated from assessments of cast and wrought Ni-based superalloys. As powder-processed Ni-based superalloys are becoming increasingly widespread, understanding the different mechanisms by which they form becomes increasingly important. Detailed characterization of MC carbides present in two experimental high Nb-content powder-processed Ni-based superalloys revealed that Hf additions affect the resultant carbide morphologies. This morphology difference was attributed to a higher magnitude of elastic strain energy along the interface associated with Hf being soluble in the MC carbide lattice. The composition of the MC carbides was studied through atom probe tomography and consisted of a complex carbonitride core, which was rich in Nb and with slight Hf segregation, surrounded by an Nb carbide shell. The characterization results of the segregation behavior of Hf in the MC carbides and the subsequent influence on their morphology were compared to density functional theory calculations and found to be in good agreement, suggesting that computational modeling can successfully be used to tailor carbide features.

Identification of stacking faults in silicon carbide by polarization-resolved second harmonic generation microscopy.

Science.gov (United States)

Hristu, Radu; Stanciu, Stefan G; Tranca, Denis E; Polychroniadis, Efstathios K; Stanciu, George A

2017-07-07

Although silicon carbide is a highly promising crystalline material for a wide range of electronic devices, extended and point defects which perturb the lattice periodicity hold deep implications with respect to device reliability. There is thus a great need for developing new methods that can detect silicon carbide defects which are detrimental to device functionality. Our experiment demonstrates that polarization-resolved second harmonic generation microscopy can extend the efficiency of the "optical signature" concept as an all-optical rapid and non-destructive set of investigation methods for the differentiation between hexagonal and cubic stacking faults in silicon carbide. This technique can be used for fast and in situ characterization and optimization of growth conditions for epilayers of silicon carbide and similar materials.

Fabrication and characterization of fully ceramic microencapsulated fuels

Energy Technology Data Exchange (ETDEWEB)

Terrani, K.A., E-mail: kurt.terrani@gmail.com [Fuel Cycle and Isotopes Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Kiggans, J.O.; Katoh, Y. [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Shimoda, K. [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Montgomery, F.C.; Armstrong, B.L.; Parish, C.M. [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Hinoki, T. [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hunn, J.D. [Fuel Cycle and Isotopes Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Snead, L.L. [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

2012-07-15

The current generation of fully ceramic microencapsulated fuels, consisting of Tristructural Isotropic fuel particles embedded in a silicon carbide matrix, is fabricated by hot pressing. Matrix powder feedstock is comprised of alumina-yttria additives thoroughly mixed with silicon carbide nanopowder using polyethyleneimine as a dispersing agent. Fuel compacts are fabricated by hot pressing the powder-fuel particle mixture at a temperature of 1800-1900 Degree-Sign C using compaction pressures of 10-20 MPa. Detailed microstructural characterization of the final fuel compacts shows that oxide additives are limited in extent and are distributed uniformly at silicon carbide grain boundaries, at triple joints between silicon carbide grains, and at the fuel particle-matrix interface.

High resolution imaging of boron carbide microstructures

International Nuclear Information System (INIS)

MacKinnon, I.D.R.; Aselage, T.; Van Deusen, S.B.

1986-01-01

Two samples of boron carbide have been examined using high resolution transmission electron microscopy (HRTEM). A hot-pressed B 13 C 2 sample shows a high density of variable width twins normal to (10*1). Subtle shifts or offsets of lattice fringes along the twin plane and normal to approx.(10*5) were also observed. A B 4 C powder showed little evidence of stacking disorder in crystalline regions

A parametric thermohydraulic study an advanced pressurized light water reactor with a tight fuel rod lattice

International Nuclear Information System (INIS)

Dalle Donne, M.; Hame, W.

1982-12-01

A parametric thermohydraulic study for an Advanced Pressurized Light Water Reactor (APWR) with a tight fuel rod lattice has been performed. The APWR improves the uranium utilisation. The APWR core should be placed in a modern German PWR plant. Within this study about 200 different reactors have been calculated. The tightening of the fuel rod lattice implies a decrease of the net electrical output of the plant, which is greater for the heterogeneous reactor than for the homogeneous reactor. APWR cores mean higher core pressure drops and higher water velocities in the core region. The cores tend to be shorter and the number of fuel rods to be higher than for the PWR. At the higher fuel rod pitch to diameter ratios (p/d) the DNB limitation is more stringent than the limitation on the fuel rod linear rating given by the necessity of reflooding after a reactor accident. The contrary is true for the lower p/d ratios. Subcooled boiling in the highest rated coolant channels occurs for the most of the calculated reactors. (orig.) [de

Recent improvements and new features in the Westinghouse lattice physics codes

International Nuclear Information System (INIS)

Huria, H.C.; Buechel, R.J.

1995-01-01

Westinghouse has been using the ANC three-dimensional, two-energy-group nodal model for nuclear analysis and fuel management calculations for standard pressurized water reactor (PWR) reload design analysis since 1988. The cross sections are obtained from PHOENIX-P, a modified version of the PHOENIX lattice physics code for all square-assembly PWR cores. The PHOENIX-H code was developed for modeling both the VVER-1000 and VVER-440 fuel lattice configurations. The PHOENIX-H code has evolved from PHOENIX-P, the primary difference being in the neutronic solution modules. The PHOENIX-P code determines the assembly flux distribution using integral transport theory-based pin-cell nodal coupling followed by two-dimensional discrete ordinates solution in x-y geometry. The PHOENIX-H code uses the two-dimensional heterogeneous response method. The other infrastructure is identical in both the codes, and they share the same 42-group cross-section library

Overview of chemical characterization of FBTR fuel

International Nuclear Information System (INIS)

Venkatesan, V.; Nandi, C.; Patil, A.B.; Prakash, Amrit; Khan, K.B.; Arun Kumar

2015-01-01

Uranium Plutonium mixed carbide fuel is the driver fuel for Fast Breeder Test Reactor (FBTR) at IGCAR. The fuel is being fabricated at Radiometallurgy Division, BARC by conventional powder metallurgy route. During the fabrication of fuel, chemical quality control of process intermediates is very important to reach stringent specification of the final fuel product. Different steps are involved in the fabrication of uranium-plutonium carbide (MC) for FBTR. The main steps in the fabrication of MC fuel pellets are carbothermic reduction (CR) of mixture of uranium oxide, plutonium oxide and graphite powder to prepare MC clinkers, crushing and milling of MC clinkers and consolidation of MC powders into fuel pellets and sintering. As a part of process control, analysis of uranium (U), plutonium (Pu), carbon in oxide graphite mixture and U, Pu, carbon, oxygen, nitrogen, MC, M 2 C 3 contents in mixed carbide powder (MC clinkers) are carried out at our laboratory. Analysis of U, Pu, carbon, oxygen, nitrogen, MC and M 2 C 3 contents in mixed carbide sintered pellets are carried out as a part of quality control. This paper describes an overview of analytical instruments used during chemical quality control of mixed carbide fuel

Neutronics performances study of silicon carbide as an inert matrix to achieve very high burn-up for light water reactor fuels

International Nuclear Information System (INIS)

Chabert, C.; Coulon-Picard, E.; Pelletier, M.

2007-01-01

In order to extend the actual limits of light water reactors, the Cea has put emphasis on the exploration of major fuel innovations that would allow us to increase the competitiveness, the safety and flexibility, while keeping the standard PWR environment. Different fuel concepts have been chosen and are actually studied to evaluate their advantages and drawbacks. The objectives of these new fuels are to increase the safety performances and to achieve a very high burn-up. One concept is a CERCER fuel with silicon carbide (SiC) as an inert matrix devoted to reduce the fuel temperature at nominal conditions. Besides the investigation of the neutronic performance, analyses on the thermomechanical performances, the fuel fabrication, the fuel reprocessing and economic aspects have been performed. This paper presents particularly neutronic results obtained for the CERCER fuel. The results show that a very high burn-up, a high safety performance and a better competitiveness cannot be achieved with this fuel concept. (authors)

Numerical simulation of direct methanol fuel cells using lattice Boltzmann method

Energy Technology Data Exchange (ETDEWEB)

Delavar, Mojtaba Aghajani; Farhadi, Mousa; Sedighi, Kurosh [Faculty of Mechanical Engineering, Babol University of Technology, Babol, P.O. Box 484 (Iran)

2010-09-15

In this study Lattice Boltzmann Method (LBM) as an alternative of conventional computational fluid dynamics method is used to simulate Direct Methanol Fuel Cell (DMFC). A two dimensional lattice Boltzmann model with 9 velocities, D2Q9, is used to solve the problem. The computational domain includes all seven parts of DMFC: anode channel, catalyst and diffusion layers, membrane and cathode channel, catalyst and diffusion layers. The model has been used to predict the flow pattern and concentration fields of different species in both clear and porous channels to investigate cell performance. The results have been compared well with results in literature for flow in porous and clear channels and cell polarization curves of the DMFC at different flow speeds and feed methanol concentrations. (author)

CFD Analysis of Square Flow Channel in Thermal Engine Rocket Adventurer for Space Nuclear Application

Energy Technology Data Exchange (ETDEWEB)

Nam, S. H.; Suh, K. Y. [Seoul National University, Seoul (Korea, Republic of); Kang, S. G. [PHILOSOPHIA, Inc., Seoul (Korea, Republic of)

2008-10-15

Solar system exploration relying on chemical rockets suffers from long trip time and high cost. In this regard nuclear propulsion is an attractive option for space exploration. The performance of Nuclear Thermal Rocket (NTR) is more than twice that of the best chemical rocket. Resorting to the pure hydrogen (H{sub 2}) propellant the NTRs can possibly achieve as high as 1,000 s of specific impulse (I{sub sp}) representing the ratio of the thrust over the fuel consumption rate, as compared to only 425 s of H{sub 2}/O{sub 2} rockets. If we reflect on the mission to Mars, NTRs would reduce the round trip time to less than 300 days, instead of over 600 days with chemical rockets. This work presents CFD analysis of one Fuel Element (FE) of Thermal Engine Rocket Adventurer (TERA). In particular, one Square Flow Channel (SFC) is analyzed in Square Lattice Honeycomb (SLHC) fuel to examine the effects of mass flow rate on rocket performance.

CFD Analysis of Square Flow Channel in Thermal Engine Rocket Adventurer for Space Nuclear Application

International Nuclear Information System (INIS)

Nam, S. H.; Suh, K. Y.; Kang, S. G.

2008-01-01

Solar system exploration relying on chemical rockets suffers from long trip time and high cost. In this regard nuclear propulsion is an attractive option for space exploration. The performance of Nuclear Thermal Rocket (NTR) is more than twice that of the best chemical rocket. Resorting to the pure hydrogen (H 2 ) propellant the NTRs can possibly achieve as high as 1,000 s of specific impulse (I sp ) representing the ratio of the thrust over the fuel consumption rate, as compared to only 425 s of H 2 /O 2 rockets. If we reflect on the mission to Mars, NTRs would reduce the round trip time to less than 300 days, instead of over 600 days with chemical rockets. This work presents CFD analysis of one Fuel Element (FE) of Thermal Engine Rocket Adventurer (TERA). In particular, one Square Flow Channel (SFC) is analyzed in Square Lattice Honeycomb (SLHC) fuel to examine the effects of mass flow rate on rocket performance

A review of carbide fuel corrosion for nuclear thermal propulsion applications

Energy Technology Data Exchange (ETDEWEB)

Pelaccio, D.G.; El-Genk, M.S. [Univ. of New Mexico, Albuquerque, NM (United States). Inst. for Space Nuclear Power Studies; Butt, D.P. [Los Alamos National Lab., NM (United States)

1993-12-01

At the operation conditions of interest in nuclear thermal propulsion reactors, carbide materials have been known to exhibit a number of life limiting phenomena. These include the formation of liquid, loss by vaporization, creep and corresponding gas flow restrictions, and local corrosion and fuel structure degradation due to excessive mechanical and/or thermal loading. In addition, the radiation environment in the reactor core can produce a substantial change in its local physical properties, which can produce high thermal stresses and corresponding stress fractures (cracking). Time-temperature history and cyclic operation of the nuclear reactor can also accelerate some of these processes. The University of New Mexico`s Institute for Space Nuclear Power Studies, under NASA sponsorship has recently initiated a study to model the complicated hydrogen corrosion process. In support of this effort, an extensive review of the open literature was performed, and a technical expert workshop was conducted. This paper summarizes the results of this review.

A Review of Carbide Fuel Corrosion for Nuclear Thermal Propulsion Applications

Science.gov (United States)

Pelaccio, Dennis G.; El-Genk, Mohamed S.; Butt, Darryl P.

1994-07-01

At the operation conditions of interest in nuclear thermal propulsion reactors, carbide materials have been known to exhibit a number of life limiting phenomena. These include the formation of liquid, loss by vaporization, creep and corresponding gas flow restrictions, and local corrosion and fuel structure degradation due to excessive mechanical and/or thermal loading. In addition, the radiation environment in the reactor core can produce a substantial change in its local physical properties, which can produce high thermal stresses and corresponding stress fractures (cracking). Time-temperature history and cyclic operation of the nuclear reactor can also accelerate some of these processes. The University of New Mexico's Institute for Space Nuclear Power Studies, under NASA sponsorship has recently initiated a study to model the complicated hydrogen corrosion process. In support of this effort, an extensive review of the open literature was performed, and a technical expert workshop was conducted. This paper summarizes the results of this review.

X-ray determination of mean square amplitudes of lattice oscillations in compounds with ZnS structure

International Nuclear Information System (INIS)

Deus, P.; Schneider, H.A.; Voland, U.

1980-01-01

A general method of determination of the mean square amplitudes of lattice oscillations (MSA) for crystals with sphalerite structure is described and applied to InP. The linearity of suitable functions of the measured integral BRAGG intensities of sin 2 theta/lambda 2 is used for the verification of the parameters selected for the correction of extinction and DTS. In this way the accuracy of the results is increased. The MSAs of the InP-sublattices are evaluated. According to theoretical expectations the MSAs of the P-sublattice are larger because of the greater contributions of optical phonons. (author)

Fuel assembly for a nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Ferrari, H M; Miller, D L; Tong, L S

1973-09-06

The subject of the patent is a spacer design applicable, primarily, to LWR, and especially, though not specifically PWR, fuel assemblies. The spacer consists of an egg-box type of assembly formed of interlocking pressed plates giving a square lattice whose openings accommodate fuel pins or regulating rods. The pressed plates are formed to provide pressed-out spring-like flanges which hold the fuel pins in position and guide the regulating rods. Additional pressed-out flanges ensure the correct configuration of the spacer structure. The spacer is designed to present as little resistance as possible to coolant flow.

Effect of deposition conditions on the properties of pyrolytic silicon carbide coatings for high-temperature gas-cooled reactor fuel particles

International Nuclear Information System (INIS)

Stinton, D.P.; Lackey, W.J.

1977-10-01

Silicon carbide coatings on HTGR microsphere fuel act as the barrier to contain metallic fission products. Silicon carbide coatings were applied by the decomposition of CH 3 SiCl 3 in a 13-cm-diam (5-in.) fluidized-bed coating furnace. The effects of temperature, CH 3 SiCl 3 supply rate and the H 2 :CH 3 SiCl 3 ratio on coating properties were studied. Deposition temperature was found to control coating density, whole particle crushing strength, coating efficiency, and microstructure. Coating density and microstructure were also partially determined by the H 2 :CH 3 SiCl 3 ratio. From this work, it appears that the rate at which high quality SiC can be deposited can be increased from 0.2 to 0.5 μm/min

Design report for an annular fuel element for accommodation of a carbide test bundle on the ring position of the KNK II/2 test zone

International Nuclear Information System (INIS)

Haefner, H.E.

1982-03-01

This report describes an annular oxide element with Mark II rods for accommodation of a 19-pin carbide test bundle on position 201 in the test zone of the second core of KNK II as well as its behavior during the period of operation. The ring element comprises within a driver wrapper in three rows of pins 102 fuel pins of 7.6 mm diameter and six structural rods for fixing the spark eroded spacers. The report deals with the ring element with its individual components fuel rod, bundle, wrappers, head and foot and describes methods, criteria and results concerning the design. The carbide test bundle to be accommodated by the annular carrier element will be treated in a separate report. The loadability of the annular element with its components is demonstrated by generally valid standards for strength criteria

Effective-field theory of the Ising model with three alternative layers on the honeycomb and square lattices

Energy Technology Data Exchange (ETDEWEB)

Deviren, Bayram [Institute of Science, Erciyes University, Kayseri 38039 (Turkey); Canko, Osman [Department of Physics, Erciyes University, Kayseri 38039 (Turkey); Keskin, Mustafa [Department of Physics, Erciyes University, Kayseri 38039 (Turkey)], E-mail: keskin@erciyes.edu.tr

2008-09-15

The Ising model with three alternative layers on the honeycomb and square lattices is studied by using the effective-field theory with correlations. We consider that the nearest-neighbor spins of each layer are coupled ferromagnetically and the adjacent spins of the nearest-neighbor layers are coupled either ferromagnetically or anti-ferromagnetically depending on the sign of the bilinear exchange interactions. We investigate the thermal variations of the magnetizations and present the phase diagrams. The phase diagrams contain the paramagnetic, ferromagnetic and anti-ferromagnetic phases, and the system also exhibits a tricritical behavior.

Effective-field theory of the Ising model with three alternative layers on the honeycomb and square lattices

International Nuclear Information System (INIS)

Deviren, Bayram; Canko, Osman; Keskin, Mustafa

2008-01-01

The Ising model with three alternative layers on the honeycomb and square lattices is studied by using the effective-field theory with correlations. We consider that the nearest-neighbor spins of each layer are coupled ferromagnetically and the adjacent spins of the nearest-neighbor layers are coupled either ferromagnetically or anti-ferromagnetically depending on the sign of the bilinear exchange interactions. We investigate the thermal variations of the magnetizations and present the phase diagrams. The phase diagrams contain the paramagnetic, ferromagnetic and anti-ferromagnetic phases, and the system also exhibits a tricritical behavior

The solubility of solid fission products in carbides and nitrides of uranium and plutonium: Pt.2. Solubility rules based on lattice parameter differences

International Nuclear Information System (INIS)

Benedict, U.

1977-01-01

The Relative Lattice Parameter Difference (RLPD) is defined for a solute element with respect to cubic carbides and nitrides of uranium and plutonium as solvents. Rules are given for the relationship between the solubility and the RLPD. NaCl type monocarbides with RLPD's from -10.2% to +7.8% are completely miscible with UC and PuC. NaCl type mononitrides with RLPD's from -7.5% to +8.5% are completely miscible with UN and PuN. The solubility in the sesquicarbides increases with decreasing RPLD and becomes complete in Pu 2 C 3 at RLPD = +4%, and in U 2 C 3 at RLPD approximately +1.5%. Solubilities are predicted on the basis of these rules for the cases where no experimental results are available

Highly efficient transition metal and nitrogen co-doped carbide-derived carbon electrocatalysts for anion exchange membrane fuel cells

Science.gov (United States)

Ratso, Sander; Kruusenberg, Ivar; Käärik, Maike; Kook, Mati; Puust, Laurits; Saar, Rando; Leis, Jaan; Tammeveski, Kaido

2018-01-01

The search for an efficient electrocatalyst for oxygen reduction reaction (ORR) to replace platinum in fuel cell cathode materials is one of the hottest topics in electrocatalysis. Among the many non-noble metal catalysts, metal/nitrogen/carbon composites made by pyrolysis of cheap materials are the most promising with control over the porosity and final structure of the catalyst a crucial point. In this work we show a method of producing a highly active ORR catalyst in alkaline media with a controllable porous structure using titanium carbide derived carbon as a base structure and dicyandiamide along with FeCl3 or CoCl2 as the dopants. The resulting transition metal-nitrogen co-doped carbide derived carbon (M/N/CDC) catalyst is highly efficient for ORR electrocatalysis with the activity in 0.1 M KOH approaching that of commercial 46.1 wt.% Pt/C. The catalyst materials are also investigated by scanning electron microscopy, Raman spectroscopy and X-ray photoelectron spectroscopy to characterise the changes in morphology and composition causing the raise in electrochemical activity. MEA performance of M/N/CDC cathode materials in H2/O2 alkaline membrane fuel cell is tested with the highest power density reached being 80 mW cm-2 compared to 90 mW cm-2 for Pt/C.

Nuclear reactor fuel elements

International Nuclear Information System (INIS)

Hindle, E. D.

1984-01-01

An array of rods is assembled to form a fuel element for a pressurized water reactor, the rods comprising zirconium alloy sheathed nuclear fuel pellets and containing helium. The helium gas pressure is selected for each rod so that it differs substantially from the helium gas pressure in its closest neighbors. In a preferred arrangement the rods are arranged in a square lattice and the helium gas pressure alternates between a relatively high value and a relatively low value so that each rod has as its closest neighbors up to four rods containing helium gas at the other pressure value

Nuclear reactor fuel elements

Energy Technology Data Exchange (ETDEWEB)

Hindle, E. D.

1984-10-16

An array of rods is assembled to form a fuel element for a pressurized water reactor, the rods comprising zirconium alloy sheathed nuclear fuel pellets and containing helium. The helium gas pressure is selected for each rod so that it differs substantially from the helium gas pressure in its closest neighbors. In a preferred arrangement the rods are arranged in a square lattice and the helium gas pressure alternates between a relatively high value and a relatively low value so that each rod has as its closest neighbors up to four rods containing helium gas at the other pressure value.

Conjecture on the critical frontier of the fully anisotropic homogeneous quenched bond-mixed potts ferromagnet in square lattice

International Nuclear Information System (INIS)

Tsallis, C.

1980-01-01

It is conjectured that a logarithmic provides a very accurate approximation of the yet unknown critical frontier of a fully anisotropic homogeneous quenched bond-mixed q-state Potts ferromagnet in square lattice, where the random coupling constant J is distributed according to the laws P(J) and P'(J) for 'horizontal' and 'vertical' bonds respectively. Such an equation contains as particular cases a great number of exact results as well as a few recent conjectures (which are definitively only approximate). (Author) [pt

Novel fabrication of silicon carbide based ceramics for nuclear applications

Science.gov (United States)

Singh, Abhishek Kumar

Advances in nuclear reactor technology and the use of gas-cooled fast reactors require the development of new materials that can operate at the higher temperatures expected in these systems. These materials include refractory alloys based on Nb, Zr, Ta, Mo, W, and Re; ceramics and composites such as SiC--SiCf; carbon--carbon composites; and advanced coatings. Besides the ability to handle higher expected temperatures, effective heat transfer between reactor components is necessary for improved efficiency. Improving thermal conductivity of the fuel can lower the center-line temperature and, thereby, enhance power production capabilities and reduce the risk of premature fuel pellet failure. Crystalline silicon carbide has superior characteristics as a structural material from the viewpoint of its thermal and mechanical properties, thermal shock resistance, chemical stability, and low radioactivation. Therefore, there have been many efforts to develop SiC based composites in various forms for use in advanced energy systems. In recent years, with the development of high yield preceramic precursors, the polymer infiltration and pyrolysis (PIP) method has aroused interest for the fabrication of ceramic based materials, for various applications ranging from disc brakes to nuclear reactor fuels. The pyrolysis of preceramic polymers allow new types of ceramic materials to be processed at relatively low temperatures. The raw materials are element-organic polymers whose composition and architecture can be tailored and varied. The primary focus of this study is to use a pyrolysis based process to fabricate a host of novel silicon carbide-metal carbide or oxide composites, and to synthesize new materials based on mixed-metal silicocarbides that cannot be processed using conventional techniques. Allylhydridopolycarbosilane (AHPCS), which is an organometal polymer, was used as the precursor for silicon carbide. Inert gas pyrolysis of AHPCS produces near-stoichiometric amorphous

Iron Carbides in Fischer–Tropsch Synthesis: Theoretical and Experimental Understanding in Epsilon-Iron Carbide Phase Assignment

International Nuclear Information System (INIS)

Liu, Xing-Wu; Cao, Zhi; Zhao, Shu; Gao, Rui

2017-01-01

As active phases in low-temperature Fischer–Tropsch synthesis for liquid fuel production, epsilon iron carbides are critically important industrial materials. However, the precise atomic structure of epsilon iron carbides remains unclear, leading to a half-century of debate on the phase assignment of the ε-Fe 2 C and ε’-Fe 2.2 C. Here, we resolve this decades-long question by a combining theoretical and experimental investigation to assign the phases unambiguously. First, we have investigated the equilibrium structures and thermal stabilities of ε-Fe x C, (x = 1, 2, 2.2, 3, 4, 6, 8) by first-principles calculations. We have also acquired X-ray diffraction patterns and Mössbauer spectra for these epsilon iron carbides, and compared them with the simulated results. These analyses indicate that the unit cell of ε-Fe 2 C contains only one type of chemical environment for Fe atoms, while ε’-Fe 2.2 C has six sets of chemically distinct Fe atoms.

Hardened over-coating fuel particle and manufacture of nuclear fuel using its fuel particle

International Nuclear Information System (INIS)

Yoshimuda, Hideharu.

1990-01-01

Coated-fuel particles comprise a coating layer formed by coating ceramics such as silicon carbide or zirconium carbide and carbons, etc. to a fuel core made of nuclear fuel materials. The fuel core generally includes oxide particles such as uranium, thorium and plutonium, having 400 to 600 μm of average grain size. The average grain size of the coated-fuel particle is usually from 800 to 900 μm. The thickness of the coating layer is usually from 150 to 250 μm. Matrix material comprising a powdery graphite and a thermosetting resin such as phenol resin, etc. is overcoated to the surface of the coated-fuel particle and hardened under heating to form a hardened overcoating layer to the coated-fuel particle. If such coated-fuel particles are used, cracks, etc. are less caused to the coating layer of the coated-fuel particles upon production, thereby enabling to prevent the damages to the coating layer. (T.M.)

Vortex lattice structures in YNi2B2C

International Nuclear Information System (INIS)

Yethiraj, M.; Paul, D.M.; Tomy, C.V.; Forgan, E.M.

1997-01-01

The authors observe a flux lattice with square symmetry in the superconductor YNi 2 B 2 C when the applied field is parallel to the c-axis of the crystal. A square lattice observed previously in the isostructural magnetic analog ErNi 2 B 2 C was attributed to the interaction between magnetic order in that system and the flux lattice. Since the Y-based compound does not order magnetically, it is clear that the structure of the flux lattice is unrelated to magnetic order. In fact, they show that the flux lines have a square cross-section when the applied field is parallel to the c-axis of the crystal, since the measured penetration depth along the 100 crystal direction is larger than the penetration depth along the 110 by approximately 60%. This is the likely reason for the square symmetry of the lattice. Although they find considerable disorder in the arrangement of the flux lines at 2.5T, no melting of the vortex lattice was observed

Approximate critical surface of the bond-mixed square-lattice Ising model

International Nuclear Information System (INIS)

Levy, S.V.F.; Tsallis, C.; Curado, E.M.F.

1979-09-01

The critical surface of the quenched bond-mixed square-lattice spin-1/2 first-neighbour-interaction ferromagnetic Ising model (with exchange interactions J 1 and J 2 ) has been investigated. Through renormalization group and heuristical procedures, a very accurate (error inferior to 3x10 -4 in the variables t sub(i) = th (J sub(i)/k sub(b)T)) approximate numerical proposal for all points of this surface is presented. This proposal simultaneously satisfies all the available exact results concerning the surface, namely P sub(c) = 1/2, t sub(c) = √2 - 1, both limiting slopes in these points, and t 2 = (1-t 1 )/(1+t 1 ) for p = 1/2. Furthemore an analytic approximation (namely (1 - p) 1n(1 + t 1 ) + p 1n(1 + t 2 ) =(1/2)1n 2) is also proposed. In what concerns the available exact results, it only fails in reproducing one of the two limiting slopes, where there is an error of 1% in the derivative: these facts result in an estimated error less than 10 -3 (in the t-variables) for any points in the surface. (Author) [pt

Development of a fuzzy logic method to build objective functions in optimization problems: application to BWR fuel lattice design

International Nuclear Information System (INIS)

Martin-del-Campo, C.; Francois, J.L.; Barragan, A.M.; Palomera, M.A.

2005-01-01

In this paper we develop a methodology based on the use of the Fuzzy Logic technique to build multi-objective functions to be used in optimization processes applied to in-core nuclear fuel management. As an example, we selected the problem of determining optimal radial fuel enrichment and gadolinia distributions in a typical 'Boiling Water Reactor (BWR)' fuel lattice. The methodology is based on the use of the mathematical capability of Fuzzy Logic to model nonlinear functions of arbitrary complexity. The utility of Fuzzy Logic is to map an input space into an output space, and the primary mechanism for doing this is a list of if-then statements called rules. The rules refer to variables and adjectives that describe those variables and, the Fuzzy Logic technique interprets the values in the input vectors and, based on the set of rules assigns values to the output vector. The methodology was developed for the radial optimization of a BWR lattice where the optimization algorithm employed is Tabu Search. The global objective is to find the optimal distribution of enrichments and burnable poison concentrations in a 10*10 BWR lattice. In order to do that, a fuzzy control inference system was developed using the Fuzzy Logic Toolbox of Matlab and it has been linked to the Tabu Search optimization process. Results show that Tabu Search combined with Fuzzy Logic performs very well, obtaining lattices with optimal fuel utilization. (authors)

Self-dual cluster renormalization group approach for the square lattice Ising model specific heat and magnetization

International Nuclear Information System (INIS)

Martin, H.O.; Tsallis, C.

1981-01-01

A simple renormalization group approach based on self-dual clusters is proposed for two-dimensional nearest-neighbour 1/2 - spin Ising model on the square lattice; it reproduces the exact critical point. The internal energy and the specific heat for vanishing external magnetic field, spontaneous magnetization and the thermal (Y sub(T)) and magnetic (Y sub(H)) critical exponents are calculated. The results obtained from the first four smallest cluster sizes strongly suggest the convergence towards the exact values when the cluster sizes increases. Even for the smallest cluster, where the calculation is very simple, the results are quite accurate, particularly in the neighbourhood of the critical point. (Author) [pt

Postirradiation results and evaluation of helium-bonded uranium--plutonium carbide fuel elements irradiated in EBR-II. Interim report

International Nuclear Information System (INIS)

Latimer, T.W.; Barner, J.O.; Kerrisk, J.F.; Green, J.L.

1976-02-01

An evaluation was made of the performance of 74 helium-bonded uranium-plutonium carbide fuel elements that were irradiated in EBR-II at 38-96 kW/m to 2-12 at. percent burnup. Only 38 of these elements have completed postirradiation examination. The higher failure rate found in fuel elements which contained high-density (greater than 95 percent theoretical density) fuel than those which contained low-density (77-91 percent theoretical density) fuel was attributed to the limited ability of the high-density fuel to swell into the void space provided in the fuel element. Increasing cladding thickness and original fuel-cladding gap size were both found to influence the failure rates for elements containing low-density fuel. Lower cladding strain and higher fission-gas release were found in high-burnup fuel elements having smear densities of less than 81 percent. Fission-gas release was usually less than 5 percent for high-density fuel, but increased with burnup to a maximum of 37 percent in low-density fuel. Maximum carburization in elements attaining 5-10 at. percent burnup and clad in Types 304 or 316 stainless steel and Incoloy 800 ranged from 36-80 μm and 38-52 μm, respectively. Strontium and barium were the fission products most frequently found in contact with the cladding but no penetration of the cladding by uranium, plutonium, or fission products was observed

Interlaboratory computational comparisons of critical fast test reactor pin lattices

International Nuclear Information System (INIS)

Mincey, J.F.; Kerr, H.T.; Durst, B.M.

1979-01-01

An objective of the Consolidated Fuel Reprocessing Program's (CFRP) nuclear engineering group at Oak Ridge National Laboratory (ORNL) is to ensure that chemical equipment components designed for the reprocessing of spent LMFBR fuel (among other fuel types) are safe from a criticality standpoint. As existing data are inadequate for the general validation of computational models describing mixed plutonium--uranium oxide systems with isotopic compositions typical of LMFBR fuel, a program of critical experiments has been initiated at the Battelle Pacific Northwest Laboratories (PNL). The first series of benchmark experiments consisted of five square-pitched lattices of unirradiated Fast Test Reactor (FTR) fuel moderated and reflected by light water. Calculations of these five experiments have been conducted by both ORNL/CFRP and PNL personnel with the purpose of exploring how accurately various computational models will predict k/sub eff/ values for such neutronic systems and if differences between k/sub eff/ values obtained with these different models are significant

Advanced technologies of production of cemented carbides and composite materials based on them

International Nuclear Information System (INIS)

Bondarenko, V.; Pavlotskaya, E.; Martynova, L.; Epik, I.

2001-01-01

The paper presents new technological processes of production of W, WC and (Ti, W)C powders, cemented carbides having a controlled carbon content, high-strength nonmagnetic nickel-bonded cemented carbides, cemented carbide-based composites having a wear-resistant antifriction working layer as well as processes of regeneration of cemented carbide waste. It is shown that these technological processes permit radical changes in the production of carbide powders and products of VK, TK, VN and KKhN cemented carbides. The processes of cemented carbide production become ecologically acceptable and free of carbon black, the use of cumbersome mixers is excluded, the power expenditure is reduced and the efficiency of labor increases. It becomes possible to control precisely the carbon content within a two-phase region -carbide-metal. A high wear resistance of parts of friction couples which are lubricated with water, benzine, kerosene, diesel fuel and other low-viscosity liquids, is ensured with increased strength and shock resistance. (author)

High performance nuclear fuel element

International Nuclear Information System (INIS)

Mordarski, W.J.; Zegler, S.T.

1980-01-01

A fuel-pellet composition is disclosed for use in fast breeder reactors. Uranium carbide particles are mixed with a powder of uraniumplutonium carbides having a stable microstructure. The resulting mixture is formed into fuel pellets. The pellets thus produced exhibit a relatively low propensity to swell while maintaining a high density

High surface area synthesis, electrochemical activity, and stability of tungsten carbide supported Pt during oxygen reduction in proton exchange membrane fuel cells

Science.gov (United States)

Chhina, H.; Campbell, S.; Kesler, O.

The oxidation of carbon catalyst supports to carbon dioxide gas leads to degradation in catalyst performance over time in proton exchange membrane fuel cells (PEMFCs). The electrochemical stability of Pt supported on tungsten carbide has been evaluated on a carbon-based gas diffusion layer (GDL) at 80 °C and compared to that of HiSpec 4000™ Pt/Vulcan XC-72R in 0.5 M H 2SO 4. Due to other electrochemical processes occurring on the GDL, detailed studies were also performed on a gold mesh substrate. The oxygen reduction reaction (ORR) activity was measured both before and after accelerated oxidation cycles between +0.6 V and +1.8 V vs. RHE. Tafel plots show that the ORR activity remained high even after accelerated oxidation tests for Pt/tungsten carbide, while the ORR activity was extremely poor after accelerated oxidation tests for HiSpec 4000™. In order to make high surface area tungsten carbide, three synthesis routes were investigated. Magnetron sputtering of tungsten on carbon was found to be the most promising route, but needs further optimization.

High surface area synthesis, electrochemical activity, and stability of tungsten carbide supported Pt during oxygen reduction in proton exchange membrane fuel cells

Energy Technology Data Exchange (ETDEWEB)

Chhina, H. [Automotive fuel cell corporation, 9000 Glenlyon Parkway, Burnaby, BC (Canada); Department of Mechanical and Industrial Engineering, 5 King' s College Road, University of Toronto, Toronto, Ontario (Canada); Campbell, S. [Automotive fuel cell corporation, 9000 Glenlyon Parkway, Burnaby, BC (Canada); Kesler, O. [Department of Mechanical and Industrial Engineering, 5 King' s College Road, University of Toronto, Toronto, Ontario (Canada)

2008-04-15

The oxidation of carbon catalyst supports to carbon dioxide gas leads to degradation in catalyst performance over time in proton exchange membrane fuel cells (PEMFCs). The electrochemical stability of Pt supported on tungsten carbide has been evaluated on a carbon-based gas diffusion layer (GDL) at 80 C and compared to that of HiSpec 4000 trademark Pt/Vulcan XC-72R in 0.5 M H{sub 2}SO{sub 4}. Due to other electrochemical processes occurring on the GDL, detailed studies were also performed on a gold mesh substrate. The oxygen reduction reaction (ORR) activity was measured both before and after accelerated oxidation cycles between +0.6 V and +1.8 V vs. RHE. Tafel plots show that the ORR activity remained high even after accelerated oxidation tests for Pt/tungsten carbide, while the ORR activity was extremely poor after accelerated oxidation tests for HiSpec 4000 trademark. In order to make high surface area tungsten carbide, three synthesis routes were investigated. Magnetron sputtering of tungsten on carbon was found to be the most promising route, but needs further optimization. (author)

Burnable poison management in light water reactor lattices

Energy Technology Data Exchange (ETDEWEB)

Buenemann, D; Mueller, A

1970-07-01

For a better reactivity control and power flattening as well as for an increase in dynamic stability the use of burnable poisons in light water reactors has been considered. The main goals for a burnable poison management and its technological realisation are discussed. The poison is assumed to be in the form of separate poison rods or homogeneous or inhomogeneous poisoning in the fuel rods. A new concept with a central poison rod within the fuel rod is discussed. The balance-equation for the needed concentration of burnable poisons for reactivity central as well as the problems of optimization of lumped poisons are treated in connection with the fuel lattice burnup. A first approximation for the design is found. For the calculation of the microburnup of lumped poison and fuel the special code NEUTRA has been developed. The burnup-equation can be chosen either in a simplified burnup-version with 2 pseudo fission products for each fissionable isotope or with an extended system of burnup equations to be used at sophisticated calculations. These burnup equations are coupled to S{sub N}-routines optionally for cylindrical or x-y-geometry for the proper calculation of the microscopic isotope density-, flux-, and power distributions. The theoretical predictions have been checked by means of special experiments so as to determine the accuracy of the computations. Even for a relatively long burnup of the fuel the predicted values are found to be within the experimental error in the case of lumped rods containing a cadmium-alloy or boron carbide. (auth)

Pilot production of 325 kg of uranium carbide

International Nuclear Information System (INIS)

Clozet, C.; Dessus, J.; Devillard, J.; Guibert, M.; Morlot, G.

1969-01-01

This report describes the pilot fabrication of uranium carbide rods to be mounted in bundles and assayed in two channels of the EL 4 reactor. The fabrication process includes: - elaboration of uranium carbide granules by carbothermic reduction of uranium dioxide; - electron bombardment melting and continuous casting of the granules; - machining of the raw ingots into rods of the required dimensions; finally, the rods will be piled-up to make the fuel elements. Both qualitative and quantitative results of this pilot production chain are presented and discussed. (authors) [fr

Evaluation of Aluminum-Boron Carbide Neutron Absorbing Materials for Interim Storage of Used Nuclear Fuel

International Nuclear Information System (INIS)

Wang, Lumin; Wierschke, Jonathan Brett

2015-01-01

The objective of this work was to understand the corrosion behavior of Boral® and Bortec® neutron absorbers over long-term deployment in a used nuclear fuel dry cask storage environment. Corrosion effects were accelerated by flowing humidified argon through an autoclave at temperatures up to 570°C. Test results show little corrosion of the aluminum matrix but that boron is leaching out of the samples. Initial tests performed at 400 and 570°C were hampered by reduced flow caused by the rapid build-up of solid deposits in the outlet lines. Analysis of the deposits by XRD shows that the deposits are comprised of boron trioxide and sassolite (H 3 BO 3 ). The collection of boron- containing compounds in the outlet lines indicated that boron was being released from the samples. Observation of the exposed samples using SEM and optical microscopy show the growth of new phases in the samples. These phases were most prominent in Bortec® samples exposed at 570°C. Samples of Boral® exposed at 570°C showed minimal new phase formation but showed nearly the complete loss of boron carbide particles. Boron carbide loss was also significant in Boral samples at 400°C. However, at 400°C phases similar to those found in Bortec® were observed. The rapid loss of the boron carbide particles in the Boral® is suspected to inhibit the formation of the new secondary phases. However, Material samples in an actual dry cask environment would be exposed to temperatures closer to 300°C and less water than the lowest test. The results from this study conclude that at the temperature and humidity levels present in a dry cask environment, corrosion and boron leaching will have no effect on the performance of Boral® and Bortec® to maintain criticality control.

The growth of minicircle networks on regular lattices

International Nuclear Information System (INIS)

Diao, Y; Hinson, K; Arsuaga, J

2012-01-01

The mitochondrial DNA of trypanosomes is organized into a network of topologically linked minicircles. In order to investigate how key topological properties of the network change with minicircle density, the authors introduced, in an earlier study, a mathematical model in which randomly oriented minicircles were placed on the vertices of the simple square lattice. Using this model, the authors rigorously showed that when the density of minicircles increases, percolation clusters form. For higher densities, these percolation clusters are the backbones for networks of minicircles that saturate the entire lattice. An important relevant question is whether these findings are generally true. That is, whether these results are independent of the choice of the lattices on which the model is based. In this paper, we study two additional lattices (namely the honeycomb and the triangular lattices). These regular lattices are selected because they have been proposed for trypanosomes before and after replication. We compare our findings with our earlier results on the square lattice and show that the mathematical statements derived for the square lattice can be extended to these other lattices qualitatively. This finding suggests the universality of these properties. Furthermore, we performed a numerical study which provided data that are consistent with our theoretical analysis, and showed that the effect of the choice of lattices on the key network topological characteristics is rather small. (paper)

The square lattice Ising model on the rectangle II: finite-size scaling limit

Science.gov (United States)

Hucht, Alfred

2017-06-01

Based on the results published recently (Hucht 2017 J. Phys. A: Math. Theor. 50 065201), the universal finite-size contributions to the free energy of the square lattice Ising model on the L× M rectangle, with open boundary conditions in both directions, are calculated exactly in the finite-size scaling limit L, M\\to∞ , T\\to Tc , with fixed temperature scaling variable x\\propto(T/Tc-1)M and fixed aspect ratio ρ\\propto L/M . We derive exponentially fast converging series for the related Casimir potential and Casimir force scaling functions. At the critical point T=Tc we confirm predictions from conformal field theory (Cardy and Peschel 1988 Nucl. Phys. B 300 377, Kleban and Vassileva 1991 J. Phys. A: Math. Gen. 24 3407). The presence of corners and the related corner free energy has dramatic impact on the Casimir scaling functions and leads to a logarithmic divergence of the Casimir potential scaling function at criticality.

Carbon potential measurement on some actinide carbides

International Nuclear Information System (INIS)

Anthonysamy, S.; Ananthasivan, K.; Kaliappan, I.; Chandramouli, V.; Vasudeva Rao, P.R.; Mathews, C.K.; Jacob, K.T.

1994-01-01

Uranium-Plutonium mixed carbides with a Pu/(U+Pu) ratio of 0.55 are to be used as the fuel in the Fast Breeder Test Reactor (FBTR) at Kalpakkam, India. Carburization of the stainless steel clad by this fuel is determined by its carbon potential. Because the carbon potential of this fuel composition is not available in the literature, it was measured by the methane-hydrogen gas equilibration technique. The sample was equilibrated with purified hydrogen and the equilibrium methane-to-hydrogen ratio in the gas phase was measured with a flame ionization detector. The carbon potential of the ThC-ThC 2 as well as Mo-Mo 2 C system, which is an important binary in the actinide-fission product-carbon systems, were also measured by this technique in the temperature range 973 to 1,173 K. The data for the Mo-Mo 2 C system are in agreement with values reported in the literature. The results for the ThC-ThC 2 system are different from estimated values with large uncertainty limits given in the literature. The data on (U, Pu) mixed carbides indicates the possibility of stainless steel clad attack under isothermal equilibrium conditions

Low temperature CVD deposition of silicon carbide

International Nuclear Information System (INIS)

Dariel, M.; Yeheskel, J.; Agam, S.; Edelstein, D.; Lebovits, O.; Ron, Y.

1991-04-01

The coating of graphite on silicon carbide from the gaseous phase in a hot-well, open flow reactor at 1150degC is described. This study constitutes the first part of an investigation of the process for the coating of nuclear fuel by chemical vapor deposition (CVD)

The German carbide program: Performance, experimental findings, and evaluation of irradiation results

International Nuclear Information System (INIS)

Steiner, H.; Freund, D.; Geithoff, D.

1982-09-01

In this report a synopsis of the German carbide program is presented. The program comprises the irradiation of about 100 carbide pins equipped with pelletted fuel. Most of these fuel pins were He-bonded, the sodium bonding concept taken as a back-up solution. The main design parameters such as smear and pellet density, gap size, pin diameter and wall thickness as well as the irradiation conditions were varied mostly within wide ranges. Based on a compilation of relevant pin parameters, irradiation conditions, and the results of various irradiation experiments conclusions on the optimum ranges of the main design parameters are drawn. Furthermore, some important aspects of fuel pin behaviour are discussed based on quantitative results from post irradiation examinations. (orig.) [de

An approach to box homogenisation-lattice properties

International Nuclear Information System (INIS)

Paul, O.P.K.

1978-01-01

A computer code has been developed to solve two group coupled neutron diffusion equations in x, y geometry for a lattice cell of a thermal reactor comprising an array of fuel pins (cellules) regularly spaced in a square box. The method uses finite difference approximation considering four neighbours of a mesh point and successive iteration technique. To simulate the current vanishing boundary condition at the cell boundary, the code uses an hypothesis that the thermal neutron flux increases exponentially beyond the pin cellules boundary while the fast neutron flux follows the reverse behaviour and the flux across the cell boundary follows the mirror image distribution. The code requires two group diffusion properties of pin cellules as input data and it calculates Ksub(infinity), L 2 , Lsub(infinity)sup(2), Dsub(th), and Df of the system. This code coupled with lattice pin and global calculation codes has been used for IRT - 2000 reactor and the results are quite reasonable. (author)

Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

Science.gov (United States)

Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pacoima, CA; Benander, Robert E [Pacoima, CA

2011-03-01

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

The growth mechanism of grain boundary carbide in Alloy 690

International Nuclear Information System (INIS)

Li, Hui; Xia, Shuang; Zhou, Bangxin; Peng, Jianchao

2013-01-01

The growth mechanism of grain boundary M 23 C 6 carbides in nickel base Alloy 690 after aging at 715 °C was investigated by high resolution transmission electron microscopy. The grain boundary carbides have coherent orientation relationship with only one side of the matrix. The incoherent phase interface between M 23 C 6 and matrix was curved, and did not lie on any specific crystal plane. The M 23 C 6 carbide transforms from the matrix phase directly at the incoherent interface. The flat coherent phase interface generally lies on low index crystal planes, such as (011) and (111) planes. The M 23 C 6 carbide transforms from a transition phase found at curved coherent phase interface. The transition phase has a complex hexagonal crystal structure, and has coherent orientation relationship with matrix and M 23 C 6 : (111) matrix //(0001) transition //(111) carbide , ¯ > matrix // ¯ 10> transition // ¯ > carbide . The crystal lattice constants of transition phase are c transition =√(3)×a matrix and a transition =√(6)/2×a matrix . Based on the experimental results, the growth mechanism of M 23 C 6 and the formation mechanism of transition phase are discussed. - Highlights: • A transition phase was observed at the coherent interfaces of M 23 C 6 and matrix. • The transition phase has hexagonal structure, and is coherent with matrix and M 23 C 6 . • The M 23 C 6 transforms from the matrix directly at the incoherent phase interface

Lattice strings

International Nuclear Information System (INIS)

Thorn, C.B.

1988-01-01

The possibility of studying non-perturbative effects in string theory using a world sheet lattice is discussed. The light-cone lattice string model of Giles and Thorn is studied numerically to assess the accuracy of ''coarse lattice'' approximations. For free strings a 5 by 15 lattice seems sufficient to obtain better than 10% accuracy for the bosonic string tachyon mass squared. In addition a crude lattice model simulating string like interactions is studied to find out how easily a coarse lattice calculation can pick out effects such as bound states which would qualitatively alter the spectrum of the free theory. The role of the critical dimension in obtaining a finite continuum limit is discussed. Instead of the ''gaussian'' lattice model one could use one of the vertex models, whose continuum limit is the same as a gaussian model on a torus of any radius. Indeed, any critical 2 dimensional statistical system will have a stringy continuum limit in the absence of string interactions. 8 refs., 1 fig. , 9 tabs

Method for placing fuel rods in individual cells, and device for performing this procedure

Energy Technology Data Exchange (ETDEWEB)

Jabsen, F S

1972-10-16

A lattice grid is described in which an egg box type assembly is formed by metal plates deformed to form spring loaded spacers which retain the fuel pins in their correct position. In order to be able to insert the fuel pins without causing scratches on their surfaces, which could lead to corrosion, the springs are displaced outwards by inserting and rotating a square-sectioned rod with rounded corners which when rotated acts as a cam, pressing the springs out. The springs are held in this position by inserting keys horizontally between the lattice plates, through holes for this purpose. The cam rod is then withdrawn, the fuel pins inserted, and the keys withdrawn. Hydraulic equipment for carrying out these operations for a large number of fuel rods simultaneously is also described.

Tungsten carbide encapsulated in nitrogen-doped carbon with iron/cobalt carbides electrocatalyst for oxygen reduction reaction

Energy Technology Data Exchange (ETDEWEB)

Zhang, Jie; Chen, Jinwei, E-mail: jwchen@scu.edu.cn; Jiang, Yiwu; Zhou, Feilong; Wang, Gang; Wang, Ruilin, E-mail: rl.wang@scu.edu.cn

2016-12-15

Graphical abstract: A hybrid catalyst was prepared via a quite green and simple method to achieve an one-pot synthesis of the N-doping carbon, tungsten carbides, and iron/cobalt carbides. It exhibited comparable electrocatalytic activity, higher durability and ability to methanol tolerance compared with commercial Pt/C to ORR. - Highlights: • A novel type of hybrid Fe/Co/WC@NC catalysts have been successfully synthesized. • The hybrid catalyst also exhibited better durability and methanol tolerance. • Multiple effective active sites of Fe{sub 3}C, Co{sub 3}C, WC, and NC help to improve catalytic performance. - Abstract: This work presents a type of hybrid catalyst prepared through an environmental and simple method, combining a pyrolysis of transition metal precursors, a nitrogen-containing material, and a tungsten source to achieve a one-pot synthesis of N-doping carbon, tungsten carbides, and iron/cobalt carbides (Fe/Co/WC@NC). The obtained Fe/Co/WC@NC consists of uniform Fe{sub 3}C and Co{sub 3}C nanoparticles encapsulated in graphitized carbon with surface nitrogen doping, closely wrapped around a plate-like tungsten carbide (WC) that functions as an efficient oxygen reduction reaction (ORR) catalyst. The introduction of WC is found to promote the ORR activity of Fe/Co-based carbide electrocatalysts, which is attributed to the synergistic catalysts of WC, Fe{sub 3}C, and Co{sub 3}C. Results suggest that the composite exhibits comparable electrocatalytic activity, higher durability, and ability for methanol tolerance compared with commercial Pt/C for ORR in alkaline electrolyte. These advantages make Fe/Co/WC@NC a promising ORR electrocatalyst and a cost-effective alternative to Pt/C for practical application as fuel cell.

Calculation methods for advanced concept light water reactor lattices

International Nuclear Information System (INIS)

Carmona, S.

1986-01-01

In the last few years s several advanced concepts for fuel rod lattices have been studied. Improved fuel utilization is one of the major aims in the development of new fuel rod designs and lattice modifications. By these changes s better performance in fuel economics s fuel burnup and material endurance can be achieved in the frame of the well-known basic Light Water Reactor technology. Among the new concepts involved in these studies that have attracted serious attention are lattices consisting of arrays of annular rods duplex pellet rods or tight multicells. These new designs of fuel rods and lattices present several computational problems. The treatment of resonance shielded cross sections is a crucial point in the analyses of these advanced concepts . The purpose of this study was to assess adequate approximation methods for calculating as accurately as possible, resonance shielding for these new lattices. Although detailed and exact computational methods for the evaluation of the resonance shielding in these lattices are possible, they are quite inefficient when used in lattice codes. The computer time and memory required for this kind of computations are too large to be used in an acceptable routine manner. In order to over- come these limitations and to make the analyses possible with reasonable use of computer resources s approximation methods are necessary. Usual approximation methods, for the resonance energy regions used in routine lattice computer codes, can not adequately handle the evaluation of these new fuel rod lattices. The main contribution of the present work to advanced lattice concepts is the development of an equivalence principle for the calculation of resonance shielding in the annular fuel pellet zone of duplex pellets; the duplex pellet in this treatment consists of two fuel zones with the same absorber isotope in both regions. In the transition from a single duplex rod to an infinite array of this kind of fuel rods, the similarity of the

Influence of dislocations in solid-phase crystal lattices on structure and properties of an WC-9Co alloy

International Nuclear Information System (INIS)

Grewe, H.

1976-01-01

After theoretical considerations about evaluation of degree of dislocation concentration in crystal lattices two tungsten-carbide-powders are characterized by chemical reaction behaviour. The hard metal grades produced from the two carbide powders are tested by material and tool life investigation. The tungsten carbide powder with lower level of dislocation-concentration leads to a hardmetall-alloy with an equal microstructure and with favourable properties, especially with a good toughness and with an interesting tool life. (orig.) [de

Molten fuel-coolant interaction behaviours of various fast reactor fuels (Paper No. HMT-45-87)

International Nuclear Information System (INIS)

Doshi, J.B.

1987-01-01

A parametric computational model of molten fuel-coolant interaction (MFCI) including a particle size distribution is developed and employed to analyse behaviours of various possible reactor fuels, such as oxide, carbide and metal in MFCI scenario. It is observed that while higher thermal conductivity and lower specific heat of carbide compared to oxide is responsible for higher peak pressure and work done per unit mass, the trend is not observed in the metal fuel. The reason for this is the lower operation temperature and latent heat of metallic fuel. (author). 9 refs., 1 fig

Thermal-hydraulics and neutronics studies on the FP7 CP-ESFR oxide and carbide cores

Energy Technology Data Exchange (ETDEWEB)

Ammirabile, L.; Tsige-Tamirat, H. [European Commission, JRC, Inst. for Energy, Petten (Netherlands)

2011-07-01

In the framework of the the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) two core designs that are currently being proposed for the 3600 MWth sodium-cooled reactor concept: one is based on oxide fuel and the other on carbide fuel. Using the European Safety Assessment Platform (ESAP), JRC-IE has conducted static calculation on neutronics (incl. reactivity coefficients) and thermal-hydraulic characteristics for both oxide and carbide reference cores. The quantities evaluated include: keff, coolant heat-up, void, and Doppler reactivity coefficients, axial and radial expansion reactivity coefficients, pin-by-pin calculated power profiles, average and peak channel temperatures. This paper presents the ESAP models applied in the study together with the relevant results for the oxide and carbide core. (author)

Thermal-hydraulics and neutronics studies on the FP7 CP-ESFR oxide and carbide cores

International Nuclear Information System (INIS)

Ammirabile, L.; Tsige-Tamirat, H.

2011-01-01

In the framework of the the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) two core designs that are currently being proposed for the 3600 MWth sodium-cooled reactor concept: one is based on oxide fuel and the other on carbide fuel. Using the European Safety Assessment Platform (ESAP), JRC-IE has conducted static calculation on neutronics (incl. reactivity coefficients) and thermal-hydraulic characteristics for both oxide and carbide reference cores. The quantities evaluated include: keff, coolant heat-up, void, and Doppler reactivity coefficients, axial and radial expansion reactivity coefficients, pin-by-pin calculated power profiles, average and peak channel temperatures. This paper presents the ESAP models applied in the study together with the relevant results for the oxide and carbide core. (author)

Highly Efficient Optical Pumping of Spin Defects in Silicon Carbide for Stimulated Microwave Emission

Science.gov (United States)

Fischer, M.; Sperlich, A.; Kraus, H.; Ohshima, T.; Astakhov, G. V.; Dyakonov, V.

2018-05-01

We investigate the pump efficiency of silicon-vacancy-related spins in silicon carbide. For a crystal inserted into a microwave cavity with a resonance frequency of 9.4 GHz, the spin population inversion factor of 75 with the saturation optical pump power of about 350 mW is achieved at room temperature. At cryogenic temperature, the pump efficiency drastically increases, owing to an exceptionally long spin-lattice relaxation time exceeding one minute. Based on the experimental results, we find realistic conditions under which a silicon carbide maser can operate in continuous-wave mode and serve as a quantum microwave amplifier.

Multilayer DNA Origami Packed on Hexagonal and Hybrid Lattices

DEFF Research Database (Denmark)

Ke, Yonggang; Voigt, Niels Vinther; Shih, William M.

2012-01-01

“Scaffolded DNA origami” has been proven to be a powerful and efficient approach to construct two-dimensional or three-dimensional objects with great complexity. Multilayer DNA origami has been demonstrated with helices packing along either honeycomb-lattice geometry or square-lattice geometry....... Here we report successful folding of multilayer DNA origami with helices arranged on a close-packed hexagonal lattice. This arrangement yields a higher density of helical packing and therefore higher resolution of spatial addressing than has been shown previously. We also demonstrate hybrid multilayer...... DNA origami with honeycomb-lattice, square-lattice, and hexagonal-lattice packing of helices all in one design. The availability of hexagonal close-packing of helices extends our ability to build complex structures using DNA nanotechnology....

Multilayer DNA origami packed on hexagonal and hybrid lattices.

Science.gov (United States)

Ke, Yonggang; Voigt, Niels V; Gothelf, Kurt V; Shih, William M

2012-01-25

"Scaffolded DNA origami" has been proven to be a powerful and efficient approach to construct two-dimensional or three-dimensional objects with great complexity. Multilayer DNA origami has been demonstrated with helices packing along either honeycomb-lattice geometry or square-lattice geometry. Here we report successful folding of multilayer DNA origami with helices arranged on a close-packed hexagonal lattice. This arrangement yields a higher density of helical packing and therefore higher resolution of spatial addressing than has been shown previously. We also demonstrate hybrid multilayer DNA origami with honeycomb-lattice, square-lattice, and hexagonal-lattice packing of helices all in one design. The availability of hexagonal close-packing of helices extends our ability to build complex structures using DNA nanotechnology. © 2011 American Chemical Society

Ultracold Atoms in a Square Lattice with Spin-Orbit Coupling: Charge Order, Superfluidity, and Topological Signatures

Science.gov (United States)

Rosenberg, Peter; Shi, Hao; Zhang, Shiwei

2017-12-01

We present an ab initio, numerically exact study of attractive fermions in square lattices with Rashba spin-orbit coupling. The ground state of this system is a supersolid, with coexisting charge and superfluid order. The superfluid is composed of both singlet and triplet pairs induced by spin-orbit coupling. We perform large-scale calculations using the auxiliary-field quantum Monte Carlo method to provide the first full, quantitative description of the charge, spin, and pairing properties of the system. In addition to characterizing the exotic physics, our results will serve as essential high-accuracy benchmarks for the intense theoretical and especially experimental efforts in ultracold atoms to realize and understand an expanding variety of quantum Hall and topological superconductor systems.

Emission of blue light from hydrogenated amorphous silicon carbide

Science.gov (United States)

Nevin, W. A.; Yamagishi, H.; Yamaguchi, M.; Tawada, Y.

1994-04-01

THE development of new electroluminescent materials is of current technological interest for use in flat-screen full-colour displays1. For such applications, amorphous inorganic semiconductors appear particularly promising, in view of the ease with which uniform films with good mechanical and electronic properties can be deposited over large areas2. Luminescence has been reported1 in the red-green part of the spectrum from amorphous silicon carbide prepared from gas-phase mixtures of silane and a carbon-containing species (usually methane or ethylene). But it is not possible to achieve blue luminescence by this approach. Here we show that the use of an aromatic species-xylene-as the source of carbon during deposition results in a form of amorphous silicon carbide that exhibits strong blue luminescence. The underlying structure of this material seems to be an unusual combination of an inorganic silicon carbide lattice with a substantial 'organic' π-conjugated carbon system, the latter dominating the emission properties. Moreover, the material can be readily doped with an electron acceptor in a manner similar to organic semiconductors3, and might therefore find applications as a conductivity- or colour-based chemical sensor.

ENTIRELY AQUEOUS SOLUTION-GEL ROUTE FOR THE PREPARATION OF ZIRCONIUM CARBIDE, HAFNIUM CARBIDE AND THEIR TERNARY CARBIDE POWDERS

Directory of Open Access Journals (Sweden)

Zhang Changrui

2016-07-01

Full Text Available An entirely aqueous solution-gel route has been developed for the synthesis of zirconium carbide, hafnium carbide and their ternary carbide powders. Zirconium oxychloride (ZrOCl₂.8H₂O, malic acid (MA and ethylene glycol (EG were dissolved in water to form the aqueous zirconium carbide precursor. Afterwards, this aqueous precursor was gelled and transformed into zirconium carbide at a relatively low temperature (1200 °C for achieving an intimate mixing of the intermediate products. Hafnium and the ternary carbide powders were also synthesized via the same aqueous route. All the zirconium, hafnium and ternary carbide powders exhibited a particle size of ∼100 nm.

Phonon spectrum, mechanical and thermophysical properties of thorium carbide

International Nuclear Information System (INIS)

Pérez Daroca, D.; Jaroszewicz, S.; Llois, A.M.; Mosca, H.O.

2013-01-01

In this work, we study, by means of density functional perturbation theory and the pseudopotential method, mechanical and thermophysical properties of thorium carbide. These properties are derived from the lattice dynamics in the quasi-harmonic approximation. The phonon spectrum of ThC presented in this article, to the best authors’ knowledge, have not been studied, neither experimentally, nor theoretically. We compare mechanical properties, volume thermal expansion and molar specific capacities with previous results and find a very good agreement

Phonon spectrum, mechanical and thermophysical properties of thorium carbide

Energy Technology Data Exchange (ETDEWEB)

Pérez Daroca, D., E-mail: pdaroca@tandar.cnea.gov.ar [Gerencia de Investigación y Aplicaciones, Comisión Nacional de Energía Atómica (Argentina); Consejo Nacional de Investigaciones Cientı´ficas y Técnicas (Argentina); Jaroszewicz, S. [Gerencia de Investigación y Aplicaciones, Comisión Nacional de Energía Atómica (Argentina); Instituto de Tecnología Jorge A. Sabato, UNSAM-CNEA (Argentina); Llois, A.M. [Gerencia de Investigación y Aplicaciones, Comisión Nacional de Energía Atómica (Argentina); Consejo Nacional de Investigaciones Cientı´ficas y Técnicas (Argentina); Mosca, H.O. [Gerencia de Investigación y Aplicaciones, Comisión Nacional de Energía Atómica (Argentina); Instituto de Tecnología Jorge A. Sabato, UNSAM-CNEA (Argentina)

2013-06-15

In this work, we study, by means of density functional perturbation theory and the pseudopotential method, mechanical and thermophysical properties of thorium carbide. These properties are derived from the lattice dynamics in the quasi-harmonic approximation. The phonon spectrum of ThC presented in this article, to the best authors’ knowledge, have not been studied, neither experimentally, nor theoretically. We compare mechanical properties, volume thermal expansion and molar specific capacities with previous results and find a very good agreement.

Irradiation and examination results of the AC-3 mixed-carbide test

International Nuclear Information System (INIS)

Mason, R.E.; Hoth, C.W.; Stratton, R.W.; Botta, F.

1992-01-01

The AC-3 test was a cooperative Swiss/US irradiation test of mixed-carbide, (U,Pr)C, fuel pins in the Fast Flux Test Facility. The test included 25 Swiss-fabricated sphere-pac-type fuel pins and 66 U.S. fabricated pellet-type fuel pins. The test was designed to operate at prototypical fast reactor conditions to provide a direct comparison of the irradiation performance of the two fuel types. The test design and fuel fabrication processes used for the AC-3 test are presented

Dependence of silicon carbide coating properties on deposition parameters: preliminary report

International Nuclear Information System (INIS)

Lauf, R.J.; Braski, D.N.

1980-05-01

Fuel particles for the High-Temperature Gas-Cooled Reactor (HTGR) contain a layer of pyrolytic silicon carbide, which acts as a pressure vessel and provides containment of metallic fission products. The silicon carbide (SiC) is deposited by the thermal decomposition of methyltrichlorosilane (CH 3 SiCl 3 or MTS) in an excess of hydrogen. The purpose of the current study is to determine how the deposition variables affect the structure and properties of the SiC layer

Technology of the production of breeder fuel elements

International Nuclear Information System (INIS)

Funke, P.

1976-01-01

A survey is presented of the fabrication of oxide and carbide fuels and of the fuel rod for fast breeders (KNK, SNR-300). The advantages of the chosen methods are explained. The main points of development concerning the oxide fuel rod are gone into. The process sequence for plutonium oxide and plutonium carbide processing is presented in a flow chart. (HR) [de

Evaluation of Aluminum-Boron Carbide Neutron Absorbing Materials for Interim Storage of Used Nuclear Fuel

Energy Technology Data Exchange (ETDEWEB)

Wang, Lumin [Univ. of Michigan, Ann Arbor, MI (United States). Department of Nuclear Engineering and Radiological Science; Wierschke, Jonathan Brett [Univ. of Michigan, Ann Arbor, MI (United States). Department of Nuclear Engineering and Radiological Science

2015-04-08

The objective of this work was to understand the corrosion behavior of Boral® and Bortec® neutron absorbers over long-term deployment in a used nuclear fuel dry cask storage environment. Corrosion effects were accelerated by flowing humidified argon through an autoclave at temperatures up to 570°C. Test results show little corrosion of the aluminum matrix but that boron is leaching out of the samples. Initial tests performed at 400 and 570°C were hampered by reduced flow caused by the rapid build-up of solid deposits in the outlet lines. Analysis of the deposits by XRD shows that the deposits are comprised of boron trioxide and sassolite (H₃BO₃). The collection of boron- containing compounds in the outlet lines indicated that boron was being released from the samples. Observation of the exposed samples using SEM and optical microscopy show the growth of new phases in the samples. These phases were most prominent in Bortec® samples exposed at 570°C. Samples of Boral® exposed at 570°C showed minimal new phase formation but showed nearly the complete loss of boron carbide particles. Boron carbide loss was also significant in Boral samples at 400°C. However, at 400°C phases similar to those found in Bortec® were observed. The rapid loss of the boron carbide particles in the Boral® is suspected to inhibit the formation of the new secondary phases. However, Material samples in an actual dry cask environment would be exposed to temperatures closer to 300°C and less water than the lowest test. The results from this study conclude that at the temperature and humidity levels present in a dry cask environment, corrosion and boron leaching will have no effect on the performance of Boral® and Bortec® to maintain criticality control.

Chemical compatibility between cladding alloys and advanced fuels

International Nuclear Information System (INIS)

Fee, D.C.; Johnson, C.E.

1975-05-01

The National Advanced Fuels Program requires chemical, mechanical, and thermophysical properties data for cladding alloys. The compatibility behavior of cladding alloys with advanced fuels is critically reviewed. in carbide fuel pins, the principal compatibility problem is cladding carburization, diffusion of carbon into the cladding matrix accompanied by carbide precipitation. Carburization changes the mechanical properties of the cladding alloy. The extent of carburization increases in sodium (versus gas) bonded fuels. The depth of carburization increases with increasing sesquicarbide (M 2 C 3 ) content of the fuel. In nitride fuel pins, the principal compatibility problem is cladding nitriding, diffusion of nitrogen into the cladding matrix accompanied by nitride precipitation. Nitriding changes the mechanical properties of the cladding alloy. In both carbide and nitride fuel pins, fission products do not migrate appreciably to the cladding and do not appear to contribute to cladding attack. 77 references. (U.S.)

Self-avoiding trails with nearest-neighbour interactions on the square lattice

International Nuclear Information System (INIS)

Bedini, A; Owczarek, A L; Prellberg, T

2013-01-01

Self-avoiding walks and self-avoiding trails, two models of a polymer coil in dilute solution, have been shown to be governed by the same universality class. On the other hand, self-avoiding walks interacting via nearest-neighbour contacts (ISAW) and self-avoiding trails interacting via multiply visited sites (ISAT) are two models of the coil-globule, or collapse transition of a polymer in dilute solution. On the square lattice it has been established numerically that the collapse transition of each model lies in a different universality class. The models differ in two substantial ways. They differ in the types of subsets of random walk configurations utilized (site self-avoidance versus bond self-avoidance) and in the type of attractive interaction. It is therefore of some interest to consider self-avoiding trails interacting via nearest-neighbour attraction (INNSAT) in order to ascertain the source of the difference in the collapse universality class. Using the flatPERM algorithm, we have performed computer simulations of this model. We present numerical evidence that the singularity in the free energy of INNSAT at the collapse transition has a similar exponent to that of the ISAW model rather than the ISAT model. This would indicate that the type of interaction used in ISAW and ISAT is the source of the difference in the universality class. (paper)

Finite-lattice form factors in free-fermion models

International Nuclear Information System (INIS)

Iorgov, N; Lisovyy, O

2011-01-01

We consider the general Z 2 -symmetric free-fermion model on the finite periodic lattice, which includes as special cases the Ising model on the square and triangular lattices and the Z n -symmetric BBS τ (2) -model with n = 2. Translating Kaufman's fermionic approach to diagonalization of Ising-like transfer matrices into the language of Grassmann integrals, we determine the transfer matrix eigenvectors and observe that they coincide with the eigenvectors of a square lattice Ising transfer matrix. This allows us to find exact finite-lattice form factors of spin operators for the statistical model and the associated finite-length quantum chains, of which the most general is equivalent to the XY chain in a transverse field

Yalina booster subcritical assembly performance with low enriched uranium fuel

International Nuclear Information System (INIS)

Talamo, Alberto; Gohar, Yousry

2011-01-01

The YALINA Booster facility is a subcritical assembly located in Minsk, Belarus. The facility has special features that result in fast and thermal neutron spectra in different zones. The fast zone of the assembly uses a lead matrix and uranium fuels with different enrichments: 90% and 36%, 36%, or 21%. The thermal zone of the assembly contains 10% enriched uranium fuel in a polyethylene matrix. This study discusses the performance of the three YALINA Booster configurations with the different fuel enrichments. In order to maintain the same subcriticality level in the three configurations, the number of fuel rods in the thermal zone is increased as the uranium fuel enrichment in the fast zone is decreased. The maximum number of fuel rods that can be loaded in the thermal zone is about 1185. Consequently, the neutron multiplication of the configuration with 21% enriched uranium fuel in the fast zone is enhanced by changing the position of the boron carbide and the natural uranium absorber rods, located between the fast and the thermal zones, to form an annular rather than a square arrangement. (author)

Yalina booster subcritical assembly performance with low enriched uranium fuel

Energy Technology Data Exchange (ETDEWEB)

Talamo, Alberto; Gohar, Yousry, E-mail: alby@anl.gov [Argonne National Laboratory, Lemont, IL (United States)

2011-07-01

The YALINA Booster facility is a subcritical assembly located in Minsk, Belarus. The facility has special features that result in fast and thermal neutron spectra in different zones. The fast zone of the assembly uses a lead matrix and uranium fuels with different enrichments: 90% and 36%, 36%, or 21%. The thermal zone of the assembly contains 10% enriched uranium fuel in a polyethylene matrix. This study discusses the performance of the three YALINA Booster configurations with the different fuel enrichments. In order to maintain the same subcriticality level in the three configurations, the number of fuel rods in the thermal zone is increased as the uranium fuel enrichment in the fast zone is decreased. The maximum number of fuel rods that can be loaded in the thermal zone is about 1185. Consequently, the neutron multiplication of the configuration with 21% enriched uranium fuel in the fast zone is enhanced by changing the position of the boron carbide and the natural uranium absorber rods, located between the fast and the thermal zones, to form an annular rather than a square arrangement. (author)

Hydrotreatment activities of supported molybdenum nitrides and carbides

Energy Technology Data Exchange (ETDEWEB)

Dolce, G.M.; Savage, P.E.; Thompson, L.T. [University of Michigan, Ann Arbor, MI (United States). Dept. of Chemical Engineering

1997-05-01

The growing need for alternative sources of transportation fuels encourages the development of new hydrotreatment catalysts. These catalysts must be active and more hydrogen efficient than the current commercial hydrotreatment catalysts. Molybdenum nitrides and carbides are attractive candidate materials possessing properties that are comparable or superior to those of commercial sulfide catalysts. This research investigated the catalytic properties of {gamma}-Al{sub 2}O{sub 3}-supported molybdenum nitrides and carbides. These catalysts were synthesized via temperature-programmed reaction of supported molybdenum oxides with ammonia or methane/hydrogen mixtures. Phase constituents and compositions were determined by X-ray diffraction, elemental analysis, and neutral activation analysis. Oxygen chemisorption was used to probe the surface properties of the catalysts. Specific activities of the molybdenum nitrides and carbides were competitive with those of a commercial sulfide catalyst for hydrodenitrogenation (HDN), hydrodesulfurization (HDS), and hydrodeoxygenation (HDO). For HDN and HDS, the catalytic activity on a molybdenum basis was a strong inverse function of the molybdenum loading. Product distributions of the HDN, HDO and HDS of a variety of heteroatom compounds indicated that several of the nitrides and carbides were more hydrogen efficient than the sulfide catalyst. 35 refs., 8 figs., 7 tabs.

Dissolution of nuclear fuel samples for analytical purposes. I

International Nuclear Information System (INIS)

Krtil, J.

1983-01-01

Main attention is devoted to procedures for dissolving fuels based on uranium metal and its alloys, uranium oxides and carbides, plutonium metal, plutonium dioxide, plutonium carbides, mixed PuC-UC carbides and mixed oxides (PuU)O 2 . Data from the literature and experience gained with the dissolution of nuclear fuel samples at the Central Control Laboratory of the Nuclear Research Institute at Rez are given. (B.S.)

Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

Science.gov (United States)

Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

2013-09-03

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

Establishment and assessment of CHF data base for square-lattice rod bundles

International Nuclear Information System (INIS)

Hwang, Dae Hyun; Seo, K. W.; Kim, K. K.; Zee, S. Q.

2002-02-01

A CHF data base is constructed for square-lattice rod bundles, and assessed with various existing CHF prediction models. The CHF data base consists of 10725 data points obtained from 147 test bundles with uniform axial power distributions and 29 test bundles with non-uniform axial power distributions. The local thermal-hydraulic conditions in the subchannels are calculated by employing a subchannel analysis code MATRA. The influence of turbulent mixing parameter on CHF is evaluated quantitatively for selected test bundles with representative cross sectional configurations. The performance of various CHF prediction models including empirical correlations for round tubes or rod bundles, theoretical DNB models such as sublayer dryout model and bubble crowding model, and CHF lookup table for round tubes, are assessed for the localized rod bundle CHF data base. In view of the analysis result, it reveals that the 1995 AECL-IPPE CHF lookup table method is one of promising models in the aspect of the prediction accuracy and the applicable range. As the result of analysis employing the CHF lookup table for 9113 data points with uniform axial heat profile, the mean and the standard deviation of P/M are calculated as 1.003 and 0.115 by HBM, 1.022 and 0.319 by DSM respectively

Contributions to the neutronic analysis of a gas-cooled fast reactor

International Nuclear Information System (INIS)

Martin-del-Campo, Cecilia; Reyes-Ramirez, Ricardo; Francois, Juan-Luis; Reinking-Cejudo, Arturo G.

2011-01-01

Highlights: → Differences on reactivity with MCNPX and TRIPOLI-4 are negligible. → Fuel lattice and core criticality calculations were done. → A higher Doppler coefficient than coolant density coefficient. → Zirconium carbide is a better reflector than silicon carbide. → Adequate active height, radial size and reflector thickness were obtained. - Abstract: In this work the Monte Carlo codes MCNPX and TRIPOLI-4 were used to perform the criticality calculations of the fuel assembly and the core configuration of a gas-cooled fast reactor (GFR) concept, currently in development. The objective is to make contributions to the neutronic analysis of a gas-cooled fast reactor. In this study the fuel assembly is based on a hexagonal lattice of fuel-pins. The materials used are uranium and plutonium carbide as fuel, silicon carbide as cladding, and helium gas as coolant. Criticality calculations were done for a fuel assembly where the axial reflector thickness was varied in order to find the optimal thickness. In order to determine the best material to be used as a reflector, in the reactor core with neutrons of high energy spectrum, criticality calculations were done for three reflector materials: zirconium carbide, silicon carbide and natural uranium. It was found that the zirconium carbide provides the best neutron reflection. Criticality calculations using different active heights were done to determine the optimal height, and the reflector thickness was adjusted. Core criticality calculations were performed with different radius sizes to determine the active radial dimension of the core. A negative temperature coefficient of reactivity was verified for the fuel. The effect on reactivity produced by changes in the coolant density was also evaluated. We present the main neutronic characteristics of a preliminary fuel and core designs for the GFR concept. ENDF-VI cross-sections libraries were used in both the MCNPX and TRIPOLI-4 codes, and we verified that the

Fine defective structure of silicon carbide powders obtained from different starting materials

Directory of Open Access Journals (Sweden)

Tomila T.V.

2006-01-01

Full Text Available The fine defective structure of silicon carbide powders obtained from silicic acid-saccharose, aerosil-saccharose, aerosil-carbon black, and hydrated cellulose-silicic acid gel systems was investigated. The relation between IR absorption characteristics and the microstructure of SiC particles obtained from different starting materials was established. The numerical relationship between the lattice parameter a and the frequency νTO is presented.

1982 Annual Status Report Plutonium Fuels and Actinide Programme

International Nuclear Information System (INIS)

Lindner, R.

1983-01-01

The programme of the Transuranium Institute has long included work on advanced fuels for fast breeder reactors. Study of the swelling of carbide and nitride fuels is now nearing completion, the retention of fission gases in bubbles of different sizes in the fuel having been quantified as function of burn-up and temperature. An important step forward has been achieved in the studies of the Equation of State of Nuclear Fuels up to 5000 K. Formation of some of the less abundant isotopes in PWR fuel has been determined experimentally. Aerosol formation during the fabrication of plutonium containing fuels, part of the activity Safe Handling of Plutonium Fuel has been studied. Head-End Processing of carbide fuels has continued experiments with high burn up mixed carbides. In the field of actinide research the preparation and characterisation of pure specimens is carried out. Effect of actinides on the properties of waste glasses is investigated

Fabrication of carbide and nitride pellets and the nitride irradiations Niloc 1 and Niloc 2

International Nuclear Information System (INIS)

Blank, H.

1991-01-01

Besides the relatively well-known advanced LMFBR mixed carbide fuel an advanced mixed nitride is also an attractive candidate for the optimised fuel cycle of the European Fast Reactor, but the present knowledge about the nitride is still insufficient and should be raised to the level of the carbide. For such an optimised fuel cycle the following general conditions have been set up for the fuel: (i) the burnup of the optimised MN and MC should be at least 15 a/o or even beyond, at moderate linear ratings of less than 75 kW/m (ii) the fuel will be used in a He-bonding pin concept and (iii) as far as available an advanced economic pellet fabrication method should be employed. (iv) The fuel structure must contain 15 - 20% porosity in order to accomodate the fission product swelling at high burnup. This report gives a comprehensive description of fuel and pellet fabrication and characterization, irradiation, and post-irradiation examination. From the results important conclusions can be drawn about future work on nitrides

Lattice Boltzmann modeling of transport phenomena in fuel cells and flow batteries

Science.gov (United States)

Xu, Ao; Shyy, Wei; Zhao, Tianshou

2017-06-01

Fuel cells and flow batteries are promising technologies to address climate change and air pollution problems. An understanding of the complex multiscale and multiphysics transport phenomena occurring in these electrochemical systems requires powerful numerical tools. Over the past decades, the lattice Boltzmann (LB) method has attracted broad interest in the computational fluid dynamics and the numerical heat transfer communities, primarily due to its kinetic nature making it appropriate for modeling complex multiphase transport phenomena. More importantly, the LB method fits well with parallel computing due to its locality feature, which is required for large-scale engineering applications. In this article, we review the LB method for gas-liquid two-phase flows, coupled fluid flow and mass transport in porous media, and particulate flows. Examples of applications are provided in fuel cells and flow batteries. Further developments of the LB method are also outlined.

Determination of lattice orientation in aluminium alloy grains by low energy gallium ion-channelling

Energy Technology Data Exchange (ETDEWEB)

Silk, Jonathan R. [Aerospace Metal Composites Ltd., RAE Road, Farnborough, GU14 6XE (United Kingdom); Dashwood, Richard J. [WMG, University of Warwick, Coventry, CV4 7AL (United Kingdom); Chater, Richard J., E-mail: r.chater@imperial.ac.u [Department of Materials, Imperial College, London SW7 2AZ (United Kingdom)

2010-06-15

Polished sections of a fine-grained aluminium, silicon carbide metal matrix composite (MMC) alloy were prepared by sputtering using a low energy gallium ion source and column (FIB). The MMC had been processed by high temperature extrusion. Images of the polished surface were recorded using the ion-induced secondary electron emission. The metal matrix grains were distinguished by gallium ion-channelling contrast from the silicon carbide component. The variation of the contrast from the aluminium grains with tilt angle can be recorded and used to determine lattice orientation with the contrast from the silicon carbide (SiC) component as a reference. This method is rapid and suits site-specific investigations where classical methods of sample preparation fail.

High order Fuchsian equations for the square lattice Ising model: χ-tilde(5)

International Nuclear Information System (INIS)

Bostan, A; Boukraa, S; Guttmann, A J; Jensen, I; Hassani, S; Zenine, N; Maillard, J-M

2009-01-01

We consider the Fuchsian linear differential equation obtained (modulo a prime) for χ-tilde (5) , the five-particle contribution to the susceptibility of the square lattice Ising model. We show that one can understand the factorization of the corresponding linear differential operator from calculations using just a single prime. A particular linear combination of χ-tilde (1) and χ-tilde (3) can be removed from χ-tilde (5) and the resulting series is annihilated by a high order globally nilpotent linear ODE. The corresponding (minimal order) linear differential operator, of order 29, splits into factors of small orders. A fifth-order linear differential operator occurs as the left-most factor of the 'depleted' differential operator and it is shown to be equivalent to the symmetric fourth power of L E , the linear differential operator corresponding to the elliptic integral E. This result generalizes what we have found for the lower order terms χ-tilde (3) and χ-tilde (4) . We conjecture that a linear differential operator equivalent to a symmetric (n - 1) th power of L E occurs as a left-most factor in the minimal order linear differential operators for all χ-tilde (n) 's

Real-Time Blood Flow Estimation Using a Recursive Least-Squares Lattice Filter

DEFF Research Database (Denmark)

Stetson, Paul F.; Jensen, Jørgen Arendt

1997-01-01

-time processing for both the periodogram and lattice-filter approaches and displays both results on a PC for comparison. Results are shown for phantom data and for demodulated data from the aorta and hepatic vein of a healthy subject. This demonstrates under clinical conditions that the lattice filter gives...

Topological phase transition in anisotropic square-octagon lattice with spin-orbit coupling and exchange field

Science.gov (United States)

Yang, Yuan; Yang, Jian; Li, Xiaobing; Zhao, Yue

2018-03-01

We investigate the topological phase transitions in an anisotropic square-octagon lattice in the presence of spin-orbit coupling and exchange field. On the basis of the Chern number and spin Chern number, we find a number of topologically distinct phases with tuning the exchange field, including time-reversal-symmetry-broken quantum spin Hall phases, quantum anomalous Hall phases and a topologically trivial phase. Particularly, we observe a coexistent state of both the quantum spin Hall effect and quantum anomalous Hall effect. Besides, by adjusting the exchange filed, we find the phase transition from time-reversal-symmetry-broken quantum spin Hall phase to spin-imbalanced and spin-polarized quantum anomalous Hall phases, providing an opportunity for quantum spin manipulation. The bulk band gap closes when topological phase transitions occur between different topological phases. Furthermore, the energy and spin spectra of the edge states corresponding to different topological phases are consistent with the topological characterization based on the Chern and spin Chern numbers.

Fuel assembly storing rack

International Nuclear Information System (INIS)

Kajimura, Haruhiko; Nakamura, Masaaki.

1997-01-01

In a nuclear fuel storage rack comprising a plurality of square-shaped rack cells arranged vertically in a lattice-like configuration, the square rack cell comprises a stainless steel having a boron content of 0.75% or less and a boron equivalent with respect to calculated thermal neutron absorbing performance of 1.0 or more, and a distance between each of the gap of the square rack cells in adjacent with each other satisfies a predetermined condition. One example of the content is that B is from 0.05 to 0.75%, Gd is from 0.05 to 1.50%, C is 0.03% or less, Si is 1.0% or less, Mn is from 0.1 to 2.0%, P is 0.03% or less, S is 0.01% or less, Cr is from 18 to 26%, Ni is from 7 to 22% and Al is 0.1% or less. The distance between the rack cells can be reduced by using a material improved with thermal neutron absorbing performance by determining the boron equivalent to a predetermined value or more and with less B-content and good fabricability. In addition, the size of the rack cell itself can be reduced. This can greatly reduce the nuclear fuel storage area. (I.S.)

Dynamic magnetic properties of the mixed spin-1 and spin-3/2 Ising system on a two-layer square lattice

International Nuclear Information System (INIS)

Temizer, Ümüt

2014-01-01

In this study, the dynamic critical behavior of the mixed spin-1 and spin-3/2 Ising system on a bilayer square lattice is studied by using the Glauber-type stochastic dynamics for both ferromagnetic/ferromagnetic (FM/FM) and antiferromagnetic/ferromagnetic (AFM/FM) interactions in the presence of a time-varying external magnetic field. The dynamic equations describing the time-dependencies of the average magnetizations are derived from the Master equation. The phases in the system are obtained by solving these dynamic equations. The temperature dependence of the dynamic magnetizations is investigated in order to characterize the nature (first- or second-order) of the dynamic phase transitions and to obtain the dynamic phase transition temperatures. The dynamic phase diagrams are constructed in seven different planes for both FM/FM and AFM/FM interactions and the effects of the related interaction parameters on the dynamic phase diagrams are examined. It is found that the dynamic phase diagrams display many dynamic critical points, such as tricritical point, triple point (TP), quadruple point (QP), double critical end point (B), multicritical point (A) and tetracritical point (M). Moreover, the reentrant behavior is observed for AFM/FM interaction in the system. - Highlights: • The mixed spin (1, 3/2) Ising system is studied on a two-layer square lattice. • The Glauber transition rates are employed to construct the dynamic equations. • The dynamic phase diagrams are presented in seven different planes. • The system displays many dynamic critical points. • The reentrant behavior is observed for AFM/FM interaction

[Study on Square Super-Lattice Pattern with Surface Discharge in Dielectric Barrier Discharge by Optical Emission Spectra].

Science.gov (United States)

Niu, Xue-jiao; Dong, Li-fang; Liu, Ying; Wang, Qian; Feng, Jian-yu

2016-02-01

Square super-lattice pattern with surface discharge consisting of central spots and dim spots is firstly observed in the mixture of argon and air by using a dielectric barrier discharge device with water electrodes. By observing the image, it is found that the central spot is located at the centriod of its surrounding four dim spots. The short-exposure image recorded by a high speed video camera shows that the dim spot results from the surface discharges (SDs). The brightness of the central spot and is quite different from that of the dim spot, which indicates that the plasma states of the central spot and the dim spot may be differentiated. The optical emission spectrum method is used to further study the several plasma parameters of the central spot and the dim spot in different argon content. The emission spectra of the N₂ second positive band (C³IIu --> B³ IIg) are measured, from which the molecule vibration temperatures of the central spot and the dim spot are calculated respectively. The broadening of spectral line 696.57 nm (2P₂-->1S₅) is used to study the electron densities of the central spot and the dim spot. It is found that the molecule vibration temperature and electron density of the dim spot are higher than those of the central spot in the same argon content The molecule vibration temperature and electron density of the central spot and the dim spot increase with the argon content increasing from 90% to 99.9%. The surface discharge induced by the volume discharge (VD) has the determinative effect on the formation of the dim spot The experimental results above play an important role in studying the formation mechanism of surface discharg&of square super-lattice pattern with surface discharge. In addition, the studies exert an influence on the application of surface discharge and volume discharge in different fields.

Reactor lattice codes

International Nuclear Information System (INIS)

Kulikowska, T.

2001-01-01

The description of reactor lattice codes is carried out on the example of the WIMSD-5B code. The WIMS code in its various version is the most recognised lattice code. It is used in all parts of the world for calculations of research and power reactors. The version WIMSD-5B is distributed free of charge by NEA Data Bank. The description of its main features given in the present lecture follows the aspects defined previously for lattice calculations in the lecture on Reactor Lattice Transport Calculations. The spatial models are described, and the approach to the energy treatment is given. Finally the specific algorithm applied in fuel depletion calculations is outlined. (author)

Tungsten carbide and tungsten-molybdenum carbides as automobile exhaust catalysts

International Nuclear Information System (INIS)

Leclercq, L.; Daubrege, F.; Gengembre, L.; Leclercq, G.; Prigent, M.

1987-01-01

Several catalyst samples of tungsten carbide and W, Mo mixed carbides with different Mo/W atom ratios, have been prepared to test their ability to remove carbon monoxide, nitric oxide and propane from a synthetic exhaust gas simulating automobile emissions. Surface characterization of the catalysts has been performed by X-ray photoelectron spectroscopy (XPS) and selective chemisorption of hydrogen and carbon monoxide. Tungsten carbide exhibits good activity for CO and NO conversion, compared to a standard three-way catalyst based on Pt and Rh. However, this W carbide is ineffective in the oxidation of propane. The Mo,W mixed carbides are markedly different having only a very low activity. 9 refs.; 10 figs.; 5 tabs

Characterization and performances of cobalt-tungsten and molybdenum-tungsten carbides as anode catalyst for PEFC

International Nuclear Information System (INIS)

Izhar, Shamsul; Yoshida, Michiko; Nagai, Masatoshi

2009-01-01

The preparation of carbon-supported cobalt-tungsten and molybdenum-tungsten carbides and their activity as an anode catalyst for a polymer electrolyte fuel cell were investigated. The electrocatalytic activity for the hydrogen oxidation reaction over the catalysts was evaluated using a single-stack fuel cell and a rotating disk electrode. The characterization of the catalysts was performed by XRD, temperature-programmed carburization, temperature-programmed reduction and X-ray photoelectron spectroscopy. The maximum power densities of the 30 wt% 873 K-carburized cobalt-tungsten and molybdenum-tungsten mixed with Ketjen carbon (cobalt-tungsten carbide (CoWC)/Ketjen black (KB) and molybdenum-tungsten carbide (MoWC)/KB) were 15.7 and 12.0 mW cm -2 , respectively, which were 14 and 11%, compared to the in-house membrane electrode assembly (MEA) prepared from a 20 wt% Pt/C catalyst. The CoWC/KB catalyst exhibited the highest maximum power density compared to the MoWC/KB and WC/KB catalysts. The 873 K-carburized CoW/KB catalyst formed the oxycarbided and/or carbided CoW that are responsible for the excellent hydrogen oxygen reaction

The 3-edge-colouring problem on the 4–8 and 3–12 lattices

International Nuclear Information System (INIS)

Fjærestad, J O

2010-01-01

We consider the problem of counting the number of 3-colourings of the edges (bonds) of the 4–8 lattice and the 3–12 lattice. These lattices are Archimedean with coordination number 3, and can be regarded as decorated versions of the square and honeycomb lattice, respectively. We solve these edge-colouring problems in the infinite-lattice limit by mapping them to other models whose solution is known. The colouring problem on the 4–8 lattice is mapped to a completely packed loop model with loop fugacity n = 3 on the square lattice, which in turn can be mapped to a 6-vertex model. The colouring problem on the 3–12 lattice is mapped to the same problem on the honeycomb lattice. The 3-edge-colouring problems on the 4–8 and 3–12 lattices are equivalent to the 3-vertex-colouring problems (and thus to the zero-temperature 3-state antiferromagnetic Potts model) on the 'square kagome' ('squagome') and 'triangular kagome' lattices, respectively

IRPHE/B and W-SS-LATTICE, Spectral Shift Reactor Lattice Experiments

International Nuclear Information System (INIS)

2003-01-01

Description: B and W has performed and analysed a series of physics experiments basically concerned with the technology of heterogeneous reactors moderated and cooled by a variable mixture of heavy and light water. A reactor so moderated is termed Spectral Shift Control Reactor (S SCR). In the practical application of this concept, the moderator mixture is rich in heavy water at the beginning of core life, so a relatively large fraction of the neutrons are epithermal and are absorbed in the fertile material. As fuel is consumed, the moderator is diluted with light water. In this way the neutron spectrum is shifted, thereby increasing the proportion of thermal neutrons and the reactivity of the system. The general objective of the S SCR Basic Physics Program was to study the nuclear properties of rod lattices moderated by D 2 O-H 2 O mixtures. The volume ratio of moderator to non-moderator in all lattices was approximately 1.0, and the fuel was either 4%-enriched UO 2 clad in stainless steel or 93%-enriched UO 2 -ThO 2 (Nth/N 15) pellets clad in aluminum. The D 2 O concentration in the moderator ranged from zero to about 90 mole %. The experimental program includes critical experiments with both types of fuel, exponential experiments at room temperature with both types of fuel, exponential experiments at elevated temperatures with the 4%-enriched UO 2 fuel, and neutron age measurements in ThO 2 lattices. The theoretical program included the development of calculation methods applicable to these systems, and the analysis and correlation of the experimental data. A first report provides the results of critical experiments performed under the Spectral Shift Control Reactor Basic Physics Program. A second report documents experimental results and theoretical interpretation of a series of twenty uniform lattice critical experiments in which the neutron spectrum is varied over a fairly broad range. A third report addresses issues that bear on the problems associated with

Determining the asymptotic buckling for the reference RB reactor lattice

International Nuclear Information System (INIS)

Martinc, R.; Sotic, O.

1969-01-01

Material buckling was measured for reference lattice of the heavy water reflected system with 2% enriched uranium fuel. Experiments were done for cores with lattice pitch values: 8, 8√2, i 16 cm. Each of these cores had heavy water reflector, as well as active reflector - heavy water lattice with natural uranium fuel. The core was reflected by natural uranium lattice in order to approach asymptotic regime in the central zone. Buckling values obtained with the natural uranium lattice as reflector are, as a rule, lower then in case of heavy water reflector [sr

The significance of strength of silicon carbide for the mechanical integrity of coated fuel particles for HTRs

International Nuclear Information System (INIS)

Bongartz, K.; Scheer, A.; Schuster, H.; Taeuber, K.

1975-01-01

Silicon carbide (SiC) and pyrocarbon are used as coating material for the HTR fuel particles. The PyC shell having a certain strength acts as a pressure vessel for the fission gases whereas the SiC shell has to retain the solid fission products in the fuel kernel. For measuring the strength of coating material the so-called Brittle Ring Test was developed. Strength and Young's modulus can be measured simultaneously with this method on SiC or PyC rings prepared out of the coating material of real fuel particles. The strength measured on the ring under a certain stress distribution which is characteristic for this method is transformed with the aid of the Weibull formalism for brittle fracture into the equivalent strength of the spherical coating shell on the fuel particle under uniform stress caused by the fission gas pressure. The values measured for the strength of the SiC were high (400-700MN/m 2 ), it could therefore be assumed that a SiC layer might contribute significantly also to the mechanical strength of the fuel coating. This assumption was confirmed by an irradiation test on coated particles with PyC-SiC-PyC coatings. There were several particles with all PyC layers broken during the irradiation, whereas the SiC layers remained intact having to withstand the fission gas pressure alone. This fact can only be explained assuming that the strength of the SiC is within the range of the values measured with the brittle ring test. The result indicates that, in optimising the coating of a fuel particle, the PyC layers of a multilayer coating should be considered alone as prospective layers for the SiC. The SiC shell, besides acting as a fission product barrier, is then also responsible for the mechanical integrity of the particle

Nuclear fuel concept for the 21st century

International Nuclear Information System (INIS)

Tulenko, J.S.; Schoessow, G.

1996-01-01

In a previous paper, the author presented his rationale for the fuel cycle for the 21st century. This cycle, driven by both environmental and economic factors, required that the fuel should be able to operate in a range from 90 000 MWd/tonne of heavy metal and above. Such an operation would require the development of a cladding material that would not undergo waterside corrosion at these ultrahigh burnups. The University of Florida is proposing a new fuel arrangement that the authors feel meets the demands of high burnup and provides a safer fuel assembly. It is believed that the liquid-metal bond concept combined with a silicon carbide composite cladding and the collapsible fission gas plenum offers outstanding potential for ultrahigh burnup fuels while providing a potentially ultrasafe reactor operation. Efforts at various facilities are under way to determine the radiation stability of silicon carbide fuel and to fabricate SiC materials that will provide the radiation stability needed. Other parameters offer strong incentives to successfully develop silicon carbide as a cladding material

Corrosion resistant cemented carbide

International Nuclear Information System (INIS)

Hong, J.

1990-01-01

This paper describes a corrosion resistant cemented carbide composite. It comprises: a granular tungsten carbide phase, a semi-continuous solid solution carbide phase extending closely adjacent at least a portion of the grains of tungsten carbide for enhancing corrosion resistance, and a substantially continuous metal binder phase. The cemented carbide composite consisting essentially of an effective amount of an anti-corrosion additive, from about 4 to about 16 percent by weight metal binder phase, and with the remaining portion being from about 84 to about 96 percent by weight metal carbide wherein the metal carbide consists essentially of from about 4 to about 30 percent by weight of a transition metal carbide or mixtures thereof selected from Group IVB and of the Periodic Table of Elements and from about 70 to about 96 percent tungsten carbide. The metal binder phase consists essentially of nickel and from about 10 to about 25 percent by weight chromium, the effective amount of an anti-corrosion additive being selected from the group consisting essentially of copper, silver, tine and combinations thereof

Search for the Heisenberg spin glass on rewired square lattices with antiferromagnetic interaction

Energy Technology Data Exchange (ETDEWEB)

Surungan, Tasrief, E-mail: tasrief@unhas.ac.id; Bansawang, B.J.; Tahir, Dahlang [Department of Physics, Hasanuddin University, Makassar, South Sulawesi 90245 (Indonesia)

2016-03-11

Spin glass (SG) is a typical magnetic system with frozen random spin orientation at low temperatures. The system exhibits rich physical properties, such as infinite number of ground states, memory effect, and aging phenomena. There are two main ingredients considered to be pivotal for the existence of SG behavior, namely, frustration and randomness. For the canonical SG system, frustration is led by the presence of competing interaction between ferromagnetic (FM) and antiferromagnetic (AF) couplings. Previously, Bartolozzi et al. [Phys. Rev. B73, 224419 (2006)], reported the SG properties of the AF Ising spins on scale free network (SFN). It is a new type of SG, different from the canonical one which requires the presence of both FM and AF couplings. In this new system, frustration is purely caused by the topological factor and its randomness is related to the irregular connectvity. Recently, Surungan et. al. [Journal of Physics: Conference Series, 640, 012001 (2015)] reported SG bahavior of AF Heisenberg model on SFN. We further investigate this type of system by studying an AF Heisenberg model on rewired square lattices. We used Replica Exchange algorithm of Monte Carlo Method and calculated the SG order parameter to search for the existence of SG phase.

Search for the Heisenberg spin glass on rewired square lattices with antiferromagnetic interaction

International Nuclear Information System (INIS)

Surungan, Tasrief; Bansawang, B.J.; Tahir, Dahlang

2016-01-01

Spin glass (SG) is a typical magnetic system with frozen random spin orientation at low temperatures. The system exhibits rich physical properties, such as infinite number of ground states, memory effect, and aging phenomena. There are two main ingredients considered to be pivotal for the existence of SG behavior, namely, frustration and randomness. For the canonical SG system, frustration is led by the presence of competing interaction between ferromagnetic (FM) and antiferromagnetic (AF) couplings. Previously, Bartolozzi et al. [Phys. Rev. B73, 224419 (2006)], reported the SG properties of the AF Ising spins on scale free network (SFN). It is a new type of SG, different from the canonical one which requires the presence of both FM and AF couplings. In this new system, frustration is purely caused by the topological factor and its randomness is related to the irregular connectvity. Recently, Surungan et. al. [Journal of Physics: Conference Series, 640, 012001 (2015)] reported SG bahavior of AF Heisenberg model on SFN. We further investigate this type of system by studying an AF Heisenberg model on rewired square lattices. We used Replica Exchange algorithm of Monte Carlo Method and calculated the SG order parameter to search for the existence of SG phase.

Critical slowing down in driven-dissipative Bose-Hubbard lattices

Science.gov (United States)

Vicentini, Filippo; Minganti, Fabrizio; Rota, Riccardo; Orso, Giuliano; Ciuti, Cristiano

2018-01-01

We explore theoretically the dynamical properties of a first-order dissipative phase transition in coherently driven Bose-Hubbard systems, describing, e.g., lattices of coupled nonlinear optical cavities. Via stochastic trajectory calculations based on the truncated Wigner approximation, we investigate the dynamical behavior as a function of system size for one-dimensional (1D) and 2D square lattices in the regime where mean-field theory predicts nonlinear bistability. We show that a critical slowing down emerges for increasing number of sites in 2D square lattices, while it is absent in 1D arrays. We characterize the peculiar properties of the collective phases in the critical region.

First-principle study of structure and stability of nickel carbides

Energy Technology Data Exchange (ETDEWEB)

Gibson, Josh S; Uddin, Jamal; Cundari, Thomas R; Bodiford, Nelli K; Wilson, Angela K [Center for Advanced Scientific Computing and Modeling, University of North Texas, Denton, TX 76203 (United States); Department of Chemistry, University of North Texas, 1155 Union Circle 305070, Denton, TX 76203 (United States)

2010-10-10

Computational studies of nickel carbides, particularly Ni{sub 2}C, are scarce. A systematic density functional theory study is reported for Ni{sub 2}C, along with NiC and Ni{sub 3}C, to understand the stability and electronic structure of nickel carbides of varying stoichiometry. A comprehensive study was executed that involved 28 trial structures of varying space group symmetry for Ni{sub 2}C. An analysis of the electronic structure, geometry and thermodynamics of Ni{sub 2}C is performed, and compared with that for Ni{sub 3}C and NiC as well as several defect structures of varying composition. It is found that the most stable ground state arrangement of Ni{sub 2}C exists within a simple orthorhombic lattice and that it has metallic character. The calculated formation energies (kcal mol{sup -1}) of NiC, Ni{sub 2}C, and Ni{sub 3}C are 48.6, 7.9 and 6.4, respectively.

Convex lattice polygons of fixed area with perimeter-dependent weights.

Science.gov (United States)

Rajesh, R; Dhar, Deepak

2005-01-01

We study fully convex polygons with a given area, and variable perimeter length on square and hexagonal lattices. We attach a weight tm to a convex polygon of perimeter m and show that the sum of weights of all polygons with a fixed area s varies as s(-theta(conv))eK(t)square root(s) for large s and t less than a critical threshold tc, where K(t) is a t-dependent constant, and theta(conv) is a critical exponent which does not change with t. Using heuristic arguments, we find that theta(conv) is 1/4 for the square lattice, but -1/4 for the hexagonal lattice. The reason for this unexpected nonuniversality of theta(conv) is traced to existence of sharp corners in the asymptotic shape of these polygons.

Characterisation of nuclear dispersion fuels. The non-destructive examination of silicon carbide by selenium immersion

Energy Technology Data Exchange (ETDEWEB)

Ambler, J.F.R.; Ferguson, I.F.

1974-07-15

The non-destructive microscopic examination of silicon-carbide-coated spheres containing uranium carbide, which involves immersing the coated spheres in selenium, is particularly suited for the examination of flaws in the coats but it is not possible to measure coating thicknesses by this method. Some coats are found to be opaque and this is related to their porosity. (auth)

The use of irradiation zones in the fuel element lattice of research piles

International Nuclear Information System (INIS)

Delattre, P.; Genthon, J.P.

1960-01-01

The first part of this report examines the advantages and disadvantages of the various types of canal which may be found in the fuel element lattice of research piles. Some examples relative to the piles EL2 and EL3 are discussed in detail. From the conclusions drawn in the last part, several norms are extracted which make it possible to define the conditions the various canals must fulfil in order that they can be used to the best possible advantage for each type of irradiation. (author) [fr

Critical heat flux tests for self-spaced square finned 7 fuel rod bundle

International Nuclear Information System (INIS)

Moon, Sang Ki; Chun, Se Young; Choi, Ki Young; Park, Jong Kuk; Hwang, Dae Hyun; Zee, Sung Quun; Kim, Keung Koo

2001-09-01

Now, KAERI is developing a new advanced reactor aimed at achieving highly enhanced safety and reliability, and improved economics. SSF (Self-Spaced Square Finned) fuel rod bundle is considered as a suitable one for the new advanced reactor. The SSF fuel rods have rectangular shapes and four fins at the corners, and are arranged in triangular geometry. While the SSF fuel rod bundle is considered to have enhanced cooling efficiency, the correlations used for commercial PWR might be able to be applied. The application results of some conventional correlations show that the SSF fuel rod bundle show an enhanced CHF performance about 10 to 40 %. When some conventional CHF correlations are applied to CHF data with a similar geometry to the SSF fuel rod bundle, conventional CHF correlations including a correlation developed in Russia are judged not to be suitable for the development of SSF fuel rod bundle and for the use in a safety analysis code. From CHF experiments for SSF 7 fuel rod bundle performed in KAERI, the following results are obtained: the CHF increases with increasing mass flux, and the CHF increasing rate decreases at high mass flux conditions. The exit quality decreases with increasing mass flux. The overall effect of the mass flux on the CHF and exit quality coincides with previous understanding. Compared to the CHF data of IPPE with the same system pressure and inlet temperature, the CHF data of KAERI show the similar values. Thus, the reliability of IPPE CHF data can be confirmed indirectly

On linear waveguides of square and triangular lattice strips: an ...

Indian Academy of Sciences (India)

Basant Lal Sharma

number of applications in science [58], some of which find applications in ... higher order of neighbour interactions in the lattice models leads to ..... an even integer. Similar ...... [25] Sharma B L 2016 Transmission of anti-plane shear waves in.

Structural stability of the square flux line lattice in YNi2B2C and LuNi2B2C studied with small angle neutron scattering

DEFF Research Database (Denmark)

Eskildsen, M.R.; Gammel, P.L.; Barber, B.P.

1997-01-01

We have studied the flux line lattice in YNi2B2C and LuNi2B2C, the nonmagnetic end members of the borocarbide superconductors using small angle neutron scattering and transport. For fields, H parallel to c, we find a square symmetric lattice which disorders rapidly above H/H-c2 similar to 0.2, well...... below the ''peak effect'' at H/H-c2 = 0.9. The results for H/H-c2 controlled by the tilt modulus c(44). For H/H-c2 > 0.2, the disordering appears to be associated with the field dependence of the shear modulus, C-66....

Evaluation of the mechanical performance of silicon carbide in TRISO fuel at high temperatures

Energy Technology Data Exchange (ETDEWEB)

Rohbeck, Nadia, E-mail: nadia.rohbeck@manchester.ac.uk; Xiao, Ping, E-mail: p.xiao@manchester.ac.uk

2016-09-15

The HTR design envisions fuel operating temperatures of up to 1000 °C and in case of an accident even 1600 °C are conceivable. To ensure safety in all conditions a thorough understanding of the impact of an extreme temperature environment is necessary. This work assesses the high temperature mechanical performance of the silicon carbide (SiC) layer within the tristructural-isotropic (TRISO) fuel particle as it poses the main barrier against fission product release into the primary circuit. Therefore, simulated fuel was fabricated by fluidized bed chemical vapour deposition; varying the deposition conditions resulted in strongly differing SiC microstructures for the various samples. Subsequently the TRISO particles were annealed in inert atmosphere at temperatures ranging from 1600 °C up to 2200 °C. Scanning electron microscopy and Raman spectroscopy showed that strong disintegration of the SiC layer occurred from 2100 °C onwards, but initial signs of porosity formation were visible already at 1800 °C. Still, the elastic modulus and hardness as measured by nanoindentation were hardly impaired. After annealing stoichiometric SiC coatings showed a reduction in fracture strength as determined by a modified crush test, however the actual annealing temperature from 1600 °C to 2000 °C had no measureable effect. Furthermore, a technique was developed to measure the elastic modulus and hardness in situ up to 500 °C using a high temperature nanoindentation facility. This approach allows conducting tests while the specimen and indenter tip are heated to a specific measurement temperature, thus obtaining reliable values for the temperature dependent mechanical properties of the material. For the SiC layer in TRISO particles it was found that the elastic modulus decreased slightly from room temperature up to 500 °C, whereas the hardness was reduced more severely to approximately half of its ambient temperature value.

Evaluation of the Mechanical Performance of Silicon Carbide in TRISO Fuel at High Temperatures

International Nuclear Information System (INIS)

Rohbeck, N.; Xiao, P.

2014-01-01

The HTR design envisions fuel operating temperatures of up to 1000°C and in case of an accident even 1600°C are conceivable. To ensure safety in all conditions a thorough understanding of the impact of an extreme temperature environment is necessary. This work assesses the high temperature mechanical performance of the silicon carbide (SiC) layer within the tristructural-isotropic (TRISO) fuel particle as it poses the main barrier against fission product release into the primary circuit. Therefore simulated fuel was fabricated by fluidized bed chemical vapour deposition; varying the deposition conditions resulted in strongly differing SiC microstructures for the various samples. Subsequently the TRISO particles were annealed in inert atmosphere at temperatures ranging from 1600°C up to 2200°C. Scanning electron microscopy and Raman spectroscopy showed that strong disintegration of the SiC layer occurred from 2100°C onwards, but initial signs of porosity formation were visible already at 1800°C. Still, the elastic modulus and hardness as measured by nanoindentation were hardly impaired. After annealing stoichiometric SiC coatings showed a reduction in fracture strength as determined by a modified crush test, however the actual annealing temperature from 1600°C to 2000°C had no measureable effect. Furthermore, a technique was developed to measure the elastic modulus and hardness in-situ up to 500°C using a high temperature nanoindentation facility. This approach allows conducting numerous tests on small sample volumes and thus promises to improve our knowledge of irradiation effects on the mechanical properties. For the SiC layer in TRISO particles it was found that the elastic modulus decreased slightly from room temperature up to 500°C, whereas the hardness was reduced more severely to approximately half of its ambient temperature value. (author)

Evaluation of the mechanical performance of silicon carbide in TRISO fuel at high temperatures

International Nuclear Information System (INIS)

Rohbeck, Nadia; Xiao, Ping

2016-01-01

The HTR design envisions fuel operating temperatures of up to 1000 °C and in case of an accident even 1600 °C are conceivable. To ensure safety in all conditions a thorough understanding of the impact of an extreme temperature environment is necessary. This work assesses the high temperature mechanical performance of the silicon carbide (SiC) layer within the tristructural-isotropic (TRISO) fuel particle as it poses the main barrier against fission product release into the primary circuit. Therefore, simulated fuel was fabricated by fluidized bed chemical vapour deposition; varying the deposition conditions resulted in strongly differing SiC microstructures for the various samples. Subsequently the TRISO particles were annealed in inert atmosphere at temperatures ranging from 1600 °C up to 2200 °C. Scanning electron microscopy and Raman spectroscopy showed that strong disintegration of the SiC layer occurred from 2100 °C onwards, but initial signs of porosity formation were visible already at 1800 °C. Still, the elastic modulus and hardness as measured by nanoindentation were hardly impaired. After annealing stoichiometric SiC coatings showed a reduction in fracture strength as determined by a modified crush test, however the actual annealing temperature from 1600 °C to 2000 °C had no measureable effect. Furthermore, a technique was developed to measure the elastic modulus and hardness in situ up to 500 °C using a high temperature nanoindentation facility. This approach allows conducting tests while the specimen and indenter tip are heated to a specific measurement temperature, thus obtaining reliable values for the temperature dependent mechanical properties of the material. For the SiC layer in TRISO particles it was found that the elastic modulus decreased slightly from room temperature up to 500 °C, whereas the hardness was reduced more severely to approximately half of its ambient temperature value.

HELIOS2: Benchmarking against experiments for hexagonal and square lattices

International Nuclear Information System (INIS)

Simeonov, T.

2009-01-01

HELIOS2, is a 2D transport theory program for fuel burnup and gamma-flux calculation. It solves the neutron and gamma transport equations in a general, two-dimensional geometry bounded by a polygon of straight lines. The applied transport solver may be chosen between: The Method of Collision Probabilities (CP) and The Method of Characteristics(MoC). The former is well known for its successful application for preparation of cross section data banks for 3D simulators for all types lattices for WWERs, PWRs, BWRs, AGRs, RBMK and CANDU reactors. The later, MoC, helps in the areas where the requirements of CP for computational power become too large of practical application. The application of HELIOS2 and The Method of Characteristics for some large from calculation point of view benchmarks is presented in this paper. The analysis combines comparisons to measured data from the Hungarian ZR-6 reactor and JAERI facility of Tank type Critical Assembly (TCA) to verify and validate HELIOS2 and MOC for WWER assembly imitators; configurations with different absorber types- ZrB 2 , B 4 C, Eu 2 O 3 and Gd 2 O 3 ; and critical configurations with stainless steel in the reflector. Core eigenvalues and reaction rates are compared. With the account for the uncertainties the results are generally excellent. Special place in this paper is given to the effect of Iron-made radial reflector. Comparisons to measurements from TIC and TCA for stainless steel and Iron reflected cores are presented. The calculated by HELIOS-2 reactivity effect is in very good agreement with the measurements. (author)

HELIOS2: Benchmarking Against Experiments for Hexagonal and Square Lattices

International Nuclear Information System (INIS)

Simeonov, T.

2009-01-01

HELIOS2, is a 2D transport theory program for fuel burnup and gamma-flux calculation. It solves the neutron and gamma transport equations in a general, two-dimensional geometry bounded by a polygon of straight lines. The applied transport solver may be chosen between: The Method of Collision Probabilities and The Method of Characteristics. The former is well known for its successful application for preparation of cross section data banks for 3D simulators for all types lattices for WWER's, PWR's, BWR's, AGR's, RBMK and CANDU reactors. The later, method of characteristics, helps in the areas where the requirements of collision probability for computational power become too large of practical application. The application of HELIOS2 and The method of characteristics for some large from calculation point of view benchmarks is presented in this paper. The analysis combines comparisons to measured data from the Hungarian ZR-6 reactor and JAERI's facility of tanktype critical assembly to verify and validate HELIOS2 and method of characteristics for WWER assembly imitators; configurations with different absorber types-ZrB2, B4C, Eu2O3 and Gd2O3; and critical configurations with stainless steel in the reflector. Core eigenvalues and reaction rates are compared. With the account for the uncertainties the results are generally excellent. Special place in this paper is given to the effect of Iron-made radial reflector. Comparisons to measurements from The Temporary International Collective and tanktype critical assembly for stainless steel and Iron reflected cores are presented. The calculated by HELIOS-2 reactivity effect is in very good agreement with the measurements. (Authors)

Method for increasing the activity of fuel cell electrodes containing tungsten carbide. Verfahren zur Steigerung der Aktivitaet von Brennstoffelektroden, die Wolframcarbid enthalten

Energy Technology Data Exchange (ETDEWEB)

Binder, H.; Koehling, A.; Kuhn, W.; Lindner, W.; Sandstede, G.

1977-10-13

An increase in the activity of electrodes containing tungsten carbide for a low-temperature fuel cell with sulfuric acid as electrolyte can be achieved, if one operates the electrodes for a few hours (5-20 h) in the presence of hydrogen and a means of reduction (formaldehyde, hydrazene) in a voltage range of between +500 and +800 mV (relative to the H/sub 2/ electrode). A corrosion resistant layer is formed, which is assumed to have the composition WC/sub X/O/sub y/H/sub z/.

A benchmark comparison of the Canadian Supercritical Water-Cooled Reactor (SCWR) 64-element fuel lattice cell parameters using various computer codes

Energy Technology Data Exchange (ETDEWEB)

Sharpe, J.; Salaun, F.; Hummel, D.; Moghrabi, A., E-mail: sharpejr@mcmaster.ca [McMaster University, Hamilton, ON (Canada); Nowak, M. [McMaster University, Hamilton, ON (Canada); Institut National Polytechnique de Grenoble, Phelma, Grenoble (France); Pencer, J. [McMaster University, Hamilton, ON (Canada); Canadian Nuclear Laboratories, Chalk River, ON, (Canada); Novog, D.; Buijs, A. [McMaster University, Hamilton, ON (Canada)

2015-07-01

Discrepancies in key lattice physics parameters have been observed between various deterministic (e.g. DRAGON and WIMS-AECL) and stochastic (MCNP, KENO) neutron transport codes in modeling previous versions of the Canadian SCWR lattice cell. Further, inconsistencies in these parameters have also been observed when using different nuclear data libraries. In this work, the predictions of k∞, various reactivity coefficients, and relative ring-averaged pin powers have been re-evaluated using these codes and libraries with the most recent 64-element fuel assembly geometry. A benchmark problem has been defined to quantify the dissimilarities between code results for a number of responses along the fuel channel under prescribed hot full power (HFP), hot zero power (HZP) and cold zero power (CZP) conditions and at several fuel burnups (0, 25 and 50 MW·d·kg{sup -1} [HM]). Results from deterministic (TRITON, DRAGON) and stochastic codes (MCNP6, KENO V.a and KENO-VI) are presented. (author)

A benchmark comparison of the Canadian Supercritical Water-Cooled Reactor (SCWR) 64-element fuel lattice cell parameters using various computer codes

International Nuclear Information System (INIS)

Sharpe, J.; Salaun, F.; Hummel, D.; Moghrabi, A.; Nowak, M.; Pencer, J.; Novog, D.; Buijs, A.

2015-01-01

Discrepancies in key lattice physics parameters have been observed between various deterministic (e.g. DRAGON and WIMS-AECL) and stochastic (MCNP, KENO) neutron transport codes in modeling previous versions of the Canadian SCWR lattice cell. Further, inconsistencies in these parameters have also been observed when using different nuclear data libraries. In this work, the predictions of k∞, various reactivity coefficients, and relative ring-averaged pin powers have been re-evaluated using these codes and libraries with the most recent 64-element fuel assembly geometry. A benchmark problem has been defined to quantify the dissimilarities between code results for a number of responses along the fuel channel under prescribed hot full power (HFP), hot zero power (HZP) and cold zero power (CZP) conditions and at several fuel burnups (0, 25 and 50 MW·d·kg"-"1 [HM]). Results from deterministic (TRITON, DRAGON) and stochastic codes (MCNP6, KENO V.a and KENO-VI) are presented. (author)

Collinear Order in Frustrated Quantum Antiferromagnet on Square Lattice (CuBr)LaNb2O7

Science.gov (United States)

Oba, Noriaki; Kageyama, Hiroshi; Kitano, Taro; Yasuda, Jun; Baba, Yoichi; Nishi, Masakazu; Hirota, Kazuma; Narumi, Yasuo; Hagiwara, Masayuki; Kindo, Koichi; Saito, Takashi; Ajiro, Yoshitami; Yoshimura, Kazuyoshi

2006-11-01

Magnetic susceptibility, heat capacity, high-field magnetization and neutron diffraction measurements have been performed on a two-dimensional S = 1/2 square-lattice system (CuBr)LaNb2O7, prepared by a topotactic ion-exchange reaction of a nonmagnetic double-layered perovskite RbLaNb2O7. (CuBr)LaNb2O7 exhibits a second-order magnetic transition at 32 K, in marked contrast to a spin-singlet nature for its Cl-based counterpart (CuCl)LaNb2O7, despite nearly identical structural parameters. The magnetic structure is a novel collinear antiferromagnetic (CAF) ordering characterized by a modulation vector q = (π, 0, π) with a reduced moment of 0.6μB. Mixed ferromagnetic nearest-neighbor (J1) and antiferromagnetic second-nearest-neighbor (J2) interactions are of comparable strength (J1/kB = -35.6 K and J2/kB = 41.3 K), placing the system in a more frustrated region of the CAF phase than ever reported.

MRT-lattice Boltzmann computations of natural convection and volumetric radiation in a tilted square enclosure

Energy Technology Data Exchange (ETDEWEB)

Moufekkir, F.; Moussaoui, M.A.; Mezrhab, A. [Laboratoire de Mecanique and Energetique, Faculte des sciences, Departement de physique 60000 Oujda (Morocco); Lemonnier, D. [Institut Pprime, CNRS-ENSMA-Univ. Poitiers, ENSMA, BP 40109, 86961 Futuroscope Chasseneuil cedex (France); Naji, H. [Universite Lille Nord de France, F-59000 Lille (France); Laboratoire Genie Civil and geo-Environnement - LGCgE- EA 4515, UArtois/FSA Bethune, F-62400 Bethune (France)

2012-04-15

A numerical analysis is carried out for natural convection while in an asymmetrically heated square cavity containing an absorbing emitting medium. The numerical approach adopted uses a hybrid thermal lattice Boltzmann method (HTLBM) in which the mass and momentum conservation equations are solved by using multiple relaxation time (MRT) model and the energy equation is solved separately by using the finite difference method (FDM). In addition, the radiative transfer equation (RTE) is treated by the discrete ordinates method (DOM) using the S8 quadrature to evaluate the source term of the energy equation. The effects of parameters such as the Rayleigh number Ra, the optical thickness {tau} and the inclination angle {phi}, are studied numerically to assess their impact on the flow and temperature distribution. The results presented in terms of isotherms, streamlines and averaged Nusselt number, show that in the absence of the radiation, the temperature and the flow fields are centro-symmetric and the cavity core is thermally stratified. However, radiation causes an overall increase in temperature and velocity gradients along both thermally active walls

Sensitivity and Uncertainty Analysis for coolant void reactivity in a CANDU Fuel Lattice Cell Model

Energy Technology Data Exchange (ETDEWEB)

Yoo, Seung Yeol; Shim, Hyung Jin [Seoul National University, Seoul (Korea, Republic of)

2016-10-15

In this study, the EPBM is implemented in Seoul National university Monte Carlo (MC) code, McCARD which has the k uncertainty evaluation capability by the adjoint-weighted perturbation (AWP) method. The implementation is verified by comparing the sensitivities of the k-eigenvalue difference to the microscopic cross sections computed by the DPBM and the direct subtractions for the TMI-1 pin-cell problem. The uncertainty of the coolant void reactivity (CVR) in a CANDU fuel lattice model due to the ENDF/B-VII.1 covariance data is calculated by its sensitivities estimated by the EPBM. The method based on the eigenvalue perturbation theory (EPBM) utilizes the 1st order adjoint-weighted perturbation (AWP) technique to estimate the sensitivity of the eigenvalue difference. Furthermore this method can be easily applied in a S/U analysis code system equipped with the eigenvalue sensitivity calculation capability. The EPBM is implemented in McCARD code and verified by showing good agreement with reference solution. Then the McCARD S/U analysis have been performed with the EPBM module for the CVR in CANDU fuel lattice problem. It shows that the uncertainty contributions of nu of {sup 235}U and gamma reaction of {sup 238}U are dominant.

Heavy water critical experiments on plutonium lattice

International Nuclear Information System (INIS)

Miyawaki, Yoshio; Shiba, Kiminori

1975-06-01

This report is the summary of physics study on plutonium lattice made in Heavy Water Critical Experiment Section of PNC. By using Deuterium Critical Assembly, physics study on plutonium lattice has been carried out since 1972. Experiments on following items were performed in a core having 22.5 cm square lattice pitch. (1) Material buckling (2) Lattice parameters (3) Local power distribution factor (4) Gross flux distribution in two region core (5) Control rod worth. Experimental results were compared with theoretical ones calculated by METHUSELAH II code. It is concluded from this study that calculation by METHUSELAH II code has acceptable accuracy in the prediction on plutonium lattice. (author)

Interim design report: fuel particle crushing

International Nuclear Information System (INIS)

Baer, J.W.; Strand, J.B.; Cook, E.J.; Miller, C.M.

1977-11-01

The double-roll fuel particle crusher was developed to fracture the silicon carbide coatings of Fort St. Vrain (FSV) fertile and fissile and large high-temperature gas-cooled reactor (LHTGR) fissile fuel particles. The report details the design task for the fuel particle crusher, including historical test information on double-roll crushers for carbide-coated fuels and the design approach selected for the cold pilot plant crusher, and shows how the design addresses the equipment goals and operational objectives. Design calculations and considerations are included to support the selection of crusher drive and gearing, the materials chosen for crushing rolls and housing, and the bearing selection. The results of the initial testing for compliance with design objectives and operational capabilities are also presented. 8 figures, 4 tables

Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

International Nuclear Information System (INIS)

Sinha, V.P.; Hegde, P.V.; Prasad, G.J.; Pal, S.; Mishra, G.P.

2012-01-01

CERMET fuel with either PuO 2 or enriched UO 2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR’s). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R and D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO 2 dispersed in uranium metal matrix pellets for three different compositions i.e. U–20 wt%UO 2 , U–25 wt%UO 2 and U–30 wt%UO 2 . It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U–UO 2 compositions.

Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

Science.gov (United States)

Sinha, V. P.; Hegde, P. V.; Prasad, G. J.; Pal, S.; Mishra, G. P.

2012-08-01

CERMET fuel with either PuO2 or enriched UO2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR's). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R & D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO2 dispersed in uranium metal matrix pellets for three different compositions i.e. U-20 wt%UO2, U-25 wt%UO2 and U-30 wt%UO2. It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U-UO2 compositions.

Fuel assembly and fuel channel box

International Nuclear Information System (INIS)

Sakuma, Toraki; Hirakawa, Hiromasa; Ishizaki, Hideaki; Nakajima, Junjiro; Aizawa, Yasuhiro.

1992-01-01

A fuel channel box has a square cylindrical shape and, in the transversal cross sectional shape, the wall thickness of a corner portion is greater than that of a central portion of the side wall except for an upper portion thereof. The upper portion of the channel box includes a region to be in contact with an upper lattice plate and a region to attach a channel spacer. Then, the wall thickness of these regions is uniform in the transversal cross section and they have the same wall thickness with that of the corner portion which has the increased wall thickness. With such a constitution, the upper portion of the channel box receives a counter force applied from the upper lattice plate upon occurrence of earthquakes and moderate it to reduce local stresses and deformation. Further, a similar region with increased wall thickness is disposed also to the lower portion of the channel box, thereby enabling to suppress the amount of coolants leaked from a portion between the lower portion and a lower tie plate, and improve the mechanical integrity of the channel box. (I.N.)

Fission-product SiC reaction in HTGR fuel

International Nuclear Information System (INIS)

Montgomery, F.

1981-01-01

The primary barrier to release of fission product from any of the fuel types into the primary circuit of the HTGR are the coatings on the fuel particles. Both pyrolytic carbon and silicon carbide coatings are very effective in retaining fission gases under normal operating conditions. One of the possible performance limitations which has been observed in irradiation tests of TRISO fuel is chemical interaction of the SiC layer with fission products. This reaction reduces the thickness of the SiC layer in TRISO particles and can lead to release of fission products from the particles if the SiC layer is completely penetrated. The experimental section of this report describes the results of work at General Atomic concerning the reaction of fission products with silicon carbide. The discussion section describes data obtained by various laboratories and includes (1) a description of the fission products which have been found to react with SiC; (2) a description of the kinetics of silicon carbide thinning caused by fission product reaction during out-of-pile thermal gradient heating and the application of these kinetics to in-pile irradiation; and (3) a comparison of silicon carbide thinning in LEU and HEU fuels

Epithermal and Thermal Spectrum Indices in Heavy Water Lattices

Energy Technology Data Exchange (ETDEWEB)

Sokolowski, E K; Jonsson, A

1967-05-15

Spectral indices have been measured by foil activation technique in a number of different D{sub 2}O-moderated lattices in the Swedish zero power reactor R0 and the pressurized exponential assembly TZ. In most cases the fuel was in the form of single rods, distributed uniformly in the lattice. Parameters in these cases were lattice pitch and fuel composition. A 31-rod cluster lattice was also investigated, with the moderator temperature varying up to 210 deg C. On the basis of these measurements, as well as measurements on cluster lattices, reported by other investigators, it has been possible to derive simple correlations for the spectral indices, which seem to be of fairly general validity for D{sub 2}O lattices. The experimental results have also been compared to calculations with the multigroup collision probability program FLEF.

Epithermal and Thermal Spectrum Indices in Heavy Water Lattices

International Nuclear Information System (INIS)

Sokolowski, E.K.; Jonsson, A.

1967-05-01

Spectral indices have been measured by foil activation technique in a number of different D 2 O-moderated lattices in the Swedish zero power reactor R0 and the pressurized exponential assembly TZ. In most cases the fuel was in the form of single rods, distributed uniformly in the lattice. Parameters in these cases were lattice pitch and fuel composition. A 31-rod cluster lattice was also investigated, with the moderator temperature varying up to 210 deg C. On the basis of these measurements, as well as measurements on cluster lattices, reported by other investigators, it has been possible to derive simple correlations for the spectral indices, which seem to be of fairly general validity for D 2 O lattices. The experimental results have also been compared to calculations with the multigroup collision probability program FLEF

Particle fuel bed tests

International Nuclear Information System (INIS)

Horn, F.L.; Powell, J.R.; Savino, J.M.

1985-01-01

Gas-cooled reactors, using packed beds of small diameter coated fuel particles have been proposed for compact, high-power systems. The particulate fuel used in the tests was 800 microns in diameter, consisting of a thoria kernel coated with 200 microns of pyrocarbon. Typically, the bed of fuel particles was contained in a ceramic cylinder with porous metallic frits at each end. A dc voltage was applied to the metallic frits and the resulting electric current heated the bed. Heat was removed by passing coolant (helium or hydrogen) through the bed. Candidate frit materials, rhenium, nickel, zirconium carbide, and zirconium oxide were unaffected, while tungsten and tungsten-rhenium lost weight and strength. Zirconium-carbide particles were tested at 2000 K in H 2 for 12 hours with no visible reaction or weight loss

Coolant Void Reactivity Analysis of CANDU Lattice

Energy Technology Data Exchange (ETDEWEB)

Park, Jin Su; Lee, Hyun Suk; Tak, Tae Woo; Lee, Deok Jung [UNIST, Ulsan (Korea, Republic of)

2016-05-15

Models of CANDU-6 and ACR-700 fuel lattices were constructed for a single bundle and 2 by 2 checkerboard to understand the physics related to CVR. Also, a familiar four factor formula was used to predict the specific contributions to reactivity change in order to achieve an understanding of the physics issues related to the CVR. At the same time, because the situation of coolant voiding should bring about a change of neutron behavior, the spectral changes and neutron current were also analyzed. The models of the CANDU- 6 and ACR-700 fuel lattices were constructed using the Monte Carlo code MCNP6 using the ENDF/B-VII.0 continuous energy cross section library based on the specification from AECL. The CANDU fuel lattice was searched through sensitivity studies of each design parameter such as fuel enrichment, fuel pitch, and types of burnable absorber for obtaining better behavior in terms of CVR. Unlike the single channel coolant voiding, the ACR-700 bundle has a positive reactivity change upon 2x2 checkerboard coolant voiding. Because of the new path for neutron moderation, the neutrons from the voided channel move to the no-void channel where they lose energy and come back to the voided channel as thermal neutrons. This phenomenon causes the positive CVR when checkerboard voiding occurs. The sensitivity study revealed the effects of the moderator to fuel volume ratio, fuel enrichment, and burnable absorber on the CVR. A fuel bundle with low moderator to fuel volume ratio and high fuel enrichment can help achieve negative CVR.

Gutzwiller variational wave function for a two-orbital Hubbard model on a square lattice

Energy Technology Data Exchange (ETDEWEB)

Muenster, Kevin Torben zu

2015-07-01

freedom of the Gutzwiller correlator. Furthermore, we discussed the implications of this parameter gauge for a more general setup. The big advantage of our diagrammatic approach lies in the fact that it simplifies decisively in the limit of infinite spatial dimensions. We obtain the exact result for Gutzwiller expectation values of single-site and two-site operators without calculating a single, non-trivial diagram. Of course, the diagrams with internal vertices contribute in finite dimensions, and their importance for phase transitions and the Fermi surface has to be studied. In the chapter 3, we therefore investigated a two-band Hubbard model on a square lattice. We introduced the Hamiltonian for two degenerate p{sub x}-p{sub y} (or d{sub xy}-d{sub yz}) orbitals where we considered electron transfers between nearest neighbors and next-nearest neighbors. The orbital degeneracy reduces the number of different hopping parameters but transitions between the two orbitals are still permitted, i.e., the local orbital quantum number is not conserved in the lattice. For two degenerate orbitals, all local Coulomb interactions can be expressed in terms of the Hubbard interaction U and the Hund's-rule coupling J. The Hubbard interaction suppresses charge fluctuations in the lattice, and the Hund's rule coupling tends to maximize the local spin. We incorporated the symmetry constraints in the Gutzwiller variational states. As our first application, we studied the ferromagnetic phase transition as a function of the model parameters for various band fillings. In general, a large density of states and a strong Hund's-rule exchange favor ferromagnetism. In the Gutzwiller wave function, the ferromagnetic order is strongly suppressed so that much larger interaction strength are needed than predicted by the Hartree-Fock solution. Moreover, the regions in parameter space where non-saturated ferromagnetism occurs are much broader in Gutzwiller theory. As shown in earlier

Ceramics as nuclear reactor fuels

International Nuclear Information System (INIS)

Reeve, K.D.

1975-01-01

Ceramics are widely accepted as nuclear reactor fuel materials, for both metal clad ceramic and all-ceramic fuel designs. Metal clad UO 2 is used commercially in large tonnages in five different power reactor designs. UO 2 pellets are made by familiar ceramic techniques but in a reactor they undergo complex thermal and chemical changes which must be thoroughly understood. Metal clad uranium-plutonium dioxide is used in present day fast breeder reactors, but may eventually be replaced by uranium-plutonium carbide or nitride. All-ceramic fuels, which are necessary for reactors operating above about 750 0 C, must incorporate one or more fission product retentive ceramic coatings. BeO-coated BeO matrix dispersion fuels and silicate glaze coated UO 2 -SiO 2 have been studied for specialised applications, but the only commercial high temperature fuel is based on graphite in which small fuel particles, each coated with vapour deposited carbon and silicon carbide, are dispersed. Ceramists have much to contribute to many aspects of fuel science and technology. (author)

Double transitions, non-Ising criticality and the critical absorbing phase in an interacting monomer–dimer model on a square lattice

International Nuclear Information System (INIS)

Nam, Keekwon; Kim, Bongsoo; Park, Sangwoong; Lee, Sung Jong

2011-01-01

We present a numerical study on an interacting monomer–dimer model with nearest neighbor repulsion on a square lattice, which possesses two symmetric absorbing states. The model is observed to exhibit two nearby continuous transitions: the Z 2 symmetry-breaking order–disorder transition and the absorbing transition with directed percolation criticality. We find that the symmetry-breaking transition shows a non-Ising critical behavior, and that the absorbing phase becomes critical, in the sense that the critical decay of the dimer density observed at the absorbing transition persists even within the absorbing phase. Our findings call for further studies on microscopic models and the corresponding continuum description belonging to the generalized voter university class. (letter)

Thermodynamic studies of thorium carbide fuel preparation and fuel-clad comptability

International Nuclear Information System (INIS)

Besmann, T.M.; Beahm, E.C.

1979-01-01

The carbothermic reduction of thorium and uranium-thorium dioxide to monocarbide has been assessed. Equilibrium calculations have yielded Th-C-O and U-Th-C-O phase equilibria and (CO) pressures generated during reduction. The (CO) pressures were found to be at least five orders of magnitude greater than any of the other 15 gaseous species considered. This confirms that the monocarbide can successfully be prepared by carbothermic reduction. The chemical compatibility of thorium carbides with the Cr-Fe-Ni content of clad alloys has been thermodynamically avaluated. Solid solutions of 5 > and 5 > and of 7 C 3 > and 7 C 3 > were the principal reaction products. The Cr-Fe-Ni content of 316 stainless steel showed much less reaction product than that for any of the other six alloys considered. (orig.) [de

Thermal-Hydraulic Aspects of Changing the Nuclear Fuel-Cladding Materials from Zircaloy to Silicon Carbides

International Nuclear Information System (INIS)

Niceno, Bojan; Pouchon, Manuel

2014-01-01

The accident in Fukushima has drastically shown the drawbacks of Zircaloy claddings despite their beneficial properties in normal use. The effect of the lack of cooling and the production of hydrogen would not have been so strong if the fuel cladding had not consisted of a zirconium (or metal) alloy. International activities have been started to search for an alternative to Zircaloy, however, still on a limited basis. A project sponsored by Swissnuclear has been conducted at Paul Scherrer Institute (PSI) with the aim to close the gap in knowledge on application of silicon carbides (SiC) as potential replacement for Zircaloys as material for nuclear fuel cladding. The work was interdisciplinary, result of collaboration between different laboratories at PSI, and has focused on SiC cladding material properties, implication of its usage on neutronics and on thermal-hydraulics. This paper summarizes thermal-hydraulic aspects of changing Zircaloy for SiC as the cladding material. The change of cladding material inevitably changes the surface properties thus making a significant impact on boiling curve, and critical heat flux (CHF). Low chemical reactivity of SiC means fewer particles in the flow (less crud), which leads to fewer failures, but also decreases the CHF. Due to differences in physical properties between SiC and Zircaloys, higher brittleness of SiC in particular, might have impact on fuel-rod assembly design, which has direct influence on flow patterns and heat transfer in the fuel assembly. Higher melting (i.e. decomposition) point for SiC means that severe accident management guidelines (SAMG) should have to be re-assessed. Not only would the core degrade later than in the case of conventional fuels, but the production of hydrogen would be quite different as well. All these issues are explored in this work in two steps; first the SiC properties which may have influence on thermal-hydraulics are outlined, then each thermal-hydraulic issues is explained from

Design and Thermal Analysis for Irradiation of Pyrolytic Carbon/Silicon Carbide Diffusion Couples in the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Gerczak, Tyler J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Smith, Kurt R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-08-01

Tristructural-isotropic (TRISO)–coated particle fuel is a promising advanced fuel concept consisting of a spherical fuel kernel made of uranium oxide and uranium carbide, surrounded by a porous carbonaceous buffer layer and successive layers of dense inner pyrolytic carbon (IPyC), silicon carbide (SiC) deposited by chemical vapor , and dense outer pyrolytic carbon (OPyC). This fuel concept is being considered for advanced reactor applications such as high temperature gas-cooled reactors (HTGRs) and molten salt reactors (MSRs), as well as for accident-tolerant fuel for light water reactors (LWRs). Development and implementation of TRISO fuel for these reactor concepts support the US Department of Energy (DOE) Office of Nuclear Energy mission to promote safe, reliable nuclear energy that is sustainable and environmentally friendly. During operation, the SiC layer serves as the primary barrier to metallic fission products and actinides not retained in the kernel. It has been observed that certain fission products are released from TRISO fuel during operation, notably, Ag, Eu, and Sr [1]. Release of these radioisotopes causes safety and maintenance concerns.

Mean-field results of the multiple-band extended Hubbard model for the square-planar CuO2 lattice

International Nuclear Information System (INIS)

Nimkar, S.; Sarma, D.D.; Krishnamurthy, H.R.; Ramasesha, S.

1993-01-01

We obtain metal-insulator phase diagrams at half-filling for the five-band extended Hubbard model of the square-planar CuO 2 lattice treated within a Hartree-Fock mean-field approximation, allowing for spiral spin-density waves. We indicate the existence of an insulating phase (covalent insulator) characterized by strong covalency effects, not identified in the earlier Zaanen-Sawatzky-Allen phase diagram. While the insulating phase is always antiferromagnetic, we also obtain an antiferromagnetic metallic phase for a certain range of interaction parameters. Performing a nonperturbative calculation of J eff , the in-plane antiferromagnetic interaction is presented as a function of the parameters in the model. We also calculate the band gap and magnetic moments at various sites and discuss critically the contrasting interpretation of the electronic structure of high-T c materials arising from photoemission and neutron-scattering experiments

Development of quantitative analytical procedures on two-phase flow in tight-lattice fuel bundles for reduced-moderation light-water reactors

International Nuclear Information System (INIS)

Ohnuki, A.; Kureta, M.; Takae, K.; Tamai, H.; Akimoto, H.; Yoshida, H.

2004-01-01

The research project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) started at Japan Atomic Energy Research Institute (JAERI) in 2002. The RMWR is a light water reactor for which a higher conversion ratio more than one can be expected. In order to attain this higher conversion ratio, triangular tight-lattice fuel bundles whose gap spacing between each fuel rod is around 1 mm are required. As for the thermal design of the RMWR core, conventional analytical methods are no good because the conventional composition equations can not predict the RMWR core with high accuracy. Then, development of new quantitative analytical procedures was carried out. Those analytical procedures are constructed by model experiments and advanced two-phase flow analysis codes. This paper describes the results of the model experiments and analytical results with the developed analysis codes. (authors)

Visible light emission from porous silicon carbide

DEFF Research Database (Denmark)

Ou, Haiyan; Lu, Weifang

2017-01-01

Light-emitting silicon carbide is emerging as an environment-friendly wavelength converter in the application of light-emitting diode based white light source for two main reasons. Firstly, SiC has very good thermal conductivity and therefore a good substrate for GaN growth in addition to the small...... lattice mismatch. Secondly, SiC material is abundant, containing no rear-earth element material as commercial phosphor. In this paper, fabrication of porous SiC is introduced, and their morphology and photoluminescence are characterized. Additionally, the carrier lifetime of the porous SiC is measured...... by time-resolved photoluminescence. The ultrashort lifetime in the order of ~70ps indicates porous SiC is very promising for the application in the ultrafast visible light communications....

On techniques of ATR lattice computation

International Nuclear Information System (INIS)

1997-08-01

Lattice computation is to compute the average nuclear constants of unit fuel lattice which are required for computing core nuclear characteristics such as core power distribution and reactivity characteristics. The main nuclear constants are infinite multiplying rate, neutron movement area, cross section for diffusion computation, local power distribution and isotope composition. As for the lattice computation code, WIMS-ATR is used, which is based on the WIMS-D code developed in U.K., and for the purpose of heightening the accuracy of analysis, which was improved by adding heavy water scattering cross section considering the temperature dependence by Honeck model. For the computation of the neutron absorption by control rods, LOIEL BLUE code is used. The extrapolation distance of neutron flux on control rod surfaces is computed by using THERMOS and DTF codes, and the lattice constants of adjoining lattices are computed by using the WIMS-ATR code. As for the WIMS-ATR code, the computation flow and nuclear data library, and as for the LOIEL BLUE code, the computation flow are explained. The local power distribution in fuel assemblies determined by the WIMS-ATR code was verified with the measured data, and the results are reported. (K.I.)

The diffusion bonding of silicon carbide and boron carbide using refractory metals

International Nuclear Information System (INIS)

Cockeram, B.V.

1999-01-01

Joining is an enabling technology for the application of structural ceramics at high temperatures. Metal foil diffusion bonding is a simple process for joining silicon carbide or boron carbide by solid-state, diffusive conversion of the metal foil into carbide and silicide compounds that produce bonding. Metal diffusion bonding trials were performed using thin foils (5 microm to 100 microm) of refractory metals (niobium, titanium, tungsten, and molybdenum) with plates of silicon carbide (both α-SiC and β-SiC) or boron carbide that were lapped flat prior to bonding. The influence of bonding temperature, bonding pressure, and foil thickness on bond quality was determined from metallographic inspection of the bonds. The microstructure and phases in the joint region of the diffusion bonds were evaluated using SEM, microprobe, and AES analysis. The use of molybdenum foil appeared to result in the highest quality bond of the metal foils evaluated for the diffusion bonding of silicon carbide and boron carbide. Bonding pressure appeared to have little influence on bond quality. The use of a thinner metal foil improved the bond quality. The microstructure of the bond region produced with either the α-SiC and β-SiC polytypes were similar

Observations of silicon carbide by high resolution transmission electron microscopy

International Nuclear Information System (INIS)

Smith, D.J.; Jepps, N.W.; Page, T.F.

1978-01-01

High resolution transmission electron microscopy techniques, principally involving direct lattice imaging, have been used as part of a study of the crystallography and phase transformation mechanics of silicon carbide polytypes. In particular, the 3C (cubic) and 6H (hexagonal) polytypes have been examined together with partially transformed structural mixtures. Although direct observation of two-dimensional atomic structures was not possible at an operating voltage of 100 kV, considerable microstructural information has been obtained by careful choice of the experimental conditions. In particular, tilted beam observations of the 0.25 nm lattice fringes have been made in the 3C polytype for two different brace 111 brace plane arrays in order to study the dimensions and coherency of finely-twinned regions together with brace 0006 brace and brace 1 0 bar1 2 brace lattice images in the 6H polytype which allow the detailed stacking operations to be resolved. Lower resolution lattice images formed with axial illumination have also been used to study the nature of the 3C → 6H transformation and results are presented showing that the transformation interface may originate with fine twinning of the 3C structure followed by growth of the resultant 6H regions. Observations have been made of the detailed stepped structure of this interface together with the stacking fault distribution in the resultant 6H material. (author)

Collinear order in frustrated quantum antiferromagnet on square lattice (CuBr)LaNb2O7

International Nuclear Information System (INIS)

Oba, Noriaki; Kageyama, Hiroshi; Kitano, Taro

2006-01-01

Magnetic susceptibility, heat capacity, high-field magnetization and neutron diffraction measurements have been performed on a two-dimensional s=1/2 square-lattice system (CuBr)LaNb 2 O 7 , prepared by a topotactic ion-exchange reaction of a nonmagnetic double-layered perovskite RbLaNb 2 O 7 . (CuBr)LaNb 2 O 7 exhibits a second-order magnetic transition at 32K, in marked contrast to a spin-singlet nature for its Cl-based counterpart (CuCl)LaNb 2 O 7 , despite nearly identical structural parameters. The magnetic structure is a novel collinear antiferromagnetic (CAF) ordering characterized by a modulation vector q=(π, 0, π) with a reduced moment of 0.6μ B . Mixed ferromagnetic nearest-neighbor (J 1 ) and antiferromagnetic second-nearest-neighbor (J 2 ) interactions are of comparable strength (J 1 /k B =-35.6K and J 2 /k B =41.3K), placing the system in a more frustrated region of the CAF phase than ever reported. (author)

Nuclear fuel assembly

International Nuclear Information System (INIS)

Wakamatsu, Mitsuo.

1974-01-01

Object: To improve a circulating flow passage of coolant so as to be able to accurately detect the temperature of coolant, rare gases contained, and the like. Structure: A fuel assembly comprising a flow regulating lattice provided with a plurality of communication holes in an axial direction, said lattice being positioned at the upper end of an outer tube in which nuclear fuel elements are received, and a neutron shielding body having a plurality of spiral coolant flow passages disposed between the lattice and the nuclear fuel elements, whereby a coolant comprised of liquid sodium or the like, which moves up passing through the coolant flow passages and the flow regulating passage, is regulated and passed through a detector mounted at the upper part of the flow regulating lattice to detect coolant temperature, flow rate, and rare gases or the like as the origin of nuclear fission contained in the coolant due to breakage of fuel elements. (Kamimura, M.)

Microstructural change and its influence on fission gas release in high burnup UO 2 fuel

Science.gov (United States)

Une, K.; Nogita, K.; Kashibe, S.; Imamura, M.

1992-06-01

The microstructural change of UO 2 fuel pellets (burnup: 6-83 GWd/t), base irradiated under LWR conditions, has been studied by detailed postirradiation examinations. The lattice parameter near the fuel rim in the irradiated UO 2 increased with burnup and appeared to become constant beyond about 50 GWd/t. This lattice dilation was mainly due to the accumulation of radiation induced point defects. Moreover, the dislocation density in the UO 2 matrix developed progressively with burnup, and eventually the tangled dislocations organized many sub-grain boundaries in the highest burnup fuel of 83 GWd/t. This sub-grain structure induced by accumulated radiation damage was compatible in appearance with SEM fractography results which revealed sub-divided grains of sub-micron size in as-fabricated grains. The influence of burnup on 85Kr release from the UO 2 fuels has been examined by means of a postirradiation annealing technique. The higher fractional release of high burnup fuels was mainly due to the burnup dependence of the fractional burst release evolved on temperature ramp. The fractional burst release was represented in terms of the square root of burnup from 6 to 83 GWd/t.

Adaptive lattice decision-feedback equalizers - Their performance and application to time-variant multipath channnels

Science.gov (United States)

Ling, F.; Proakis, J. G.

1985-04-01

This paper presents two types of adaptive lattice decision-feedback equalizers (DFE), the least squares (LS) lattice DFE and the gradient lattice DFE. Their performance has been investigated on both time-invariant and time-variant channels through computer simulations and compared to other kinds of equalizers. An analysis of the self-noise and tracking characteristics of the LS DFE and the DFE employing the Widrow-Hoff least mean square adaptive algorithm (LMS DFE) are also given. The analysis and simulation results show that the LS lattice DFE has the faster initial convergence rate, while the gradient lattice DFE is computationally more efficient. The main advantages of the lattice DFE's are their numerical stability, their computational efficiency, the flexibility to change their length, and their excellent capabilities for tracking rapidly time-variant channels.

Fuel cycles with high fuel burn-up: analysis of reactivity coefficients

International Nuclear Information System (INIS)

Kryuchkov, E.F.; Shmelev, A.N.; Ternovykh, M.J.; Tikhomirov, G.V.; Jinhong, L.; Saito, M.

2003-01-01

Fuel cycles of light-water reactors (LWR) with high fuel burn-up (above 100 MWd/kg), as a rule, involve large amounts of fissionable materials. It leads to forming the neutron spectrum harder than that in traditional LWR. Change of neutron spectrum and significant amount of non-traditional isotopes (for example, 237 Np, 238 Pu, 231 Pa, 232 U) in such fuel compositions can alter substantially reactivity coefficients as compared with traditional uranium-based fuel. The present work addresses the fuel cycles with high fuel burn-up which are based on Th-Pa-U and U-Np-Pu fuel compositions. Numerical analyses are carried out to determine effective neutron multiplication factor and void reactivity coefficient (VRC) for different values of fuel burn-up and different lattice parameters. The algorithm is proposed for analysis of isotopes contribution to these coefficients. Various ways are considered to upgrade safety of nuclear fuel cycles with high fuel burn-up. So, the results obtained in this study have demonstrated that: -1) Non-traditional fuel compositions developed for achievement of high fuel burn-up in LWR can possess positive values of reactivity coefficients that is unacceptable from the reactor operation safety point of view; -2) The lattice pitch of traditional LWR is not optimal for non-traditional fuel compositions, the increased value of the lattice pitch leads to larger value of initial reactivity margin and provides negative VRC within sufficiently broad range of coolant density; -3) Fuel burn-up has an insignificant effect on VRC dependence on coolant density, so, the measures undertaken to suppress positive VRC of fresh fuel will be effective for partially burnt-up fuel compositions also and; -4) Increase of LWR core height and introduction of additional moderators into the fuel lattice can be used as the ways to reach negative VRC values for full range of possible coolant density variations

The precipitation behavior of titanium carbide on the surface of SUS 321 stainless steel

International Nuclear Information System (INIS)

Yoshihara, Kazuhiro; Nii, Kazuyoshi

1982-01-01

The surface composition of SUS 321 stainless steel at high temperatures was observed in vacuum with Auger electron spectroscopy. The precipitation of titanium carbide was found on the surface of SUS 321. The thickness of precipitated titanium carbide layer increased in proportion to the square root of annealing time and became about 0.05 μm after heated at 1100 K for 432 ks. The precipitated titanium carbide was not replaced by the most surface active element sulfur, and remained stable on the surface. The precipitated layer, however, was not even and had many holes about 1 μm in diameter. The bottom of a hole was SUS 321, on which phosphorus, oxygen and sulfur segregated. As the annealing time was prolonged, these segregants were replaced one by one in the order of the surface activity, and finally the most surface active element, sulfur, remained on the bottom of the hole. Moreover, sulfur diffused over the outside of the hole. The precipitation of titanium carbide on the surface occurred according to the following processes: (1) The titanium and carbon which had been dissolved in the bulk diffused onto the surface of the stainless steel. (2) The titanium carbide which had been precipitated in the bulk dissolved because the concentration of titanum and carbon fell under their solubility limits in the bulk. (3) The titanium and carbon diffused onto the surface which was exposed to vacuum. (4) The titanium and carbon recombined into titanium carbide and precipitated on the surface. The growth rate of the thickness of the precipitated layer was controlled by the diffusion of titanium and carbon in the precipitated titanium carbide. (J.P.N.)

STATUS OF HIGH FLUX ISOTOPE REACTOR IRRADIATION OF SILICON CARBIDE/SILICON CARBIDE JOINTS

Energy Technology Data Exchange (ETDEWEB)

Katoh, Yutai [ORNL; Koyanagi, Takaaki [ORNL; Kiggans, Jim [ORNL; Cetiner, Nesrin [ORNL; McDuffee, Joel [ORNL

2014-09-01

Development of silicon carbide (SiC) joints that retain adequate structural and functional properties in the anticipated service conditions is a critical milestone toward establishment of advanced SiC composite technology for the accident-tolerant light water reactor (LWR) fuels and core structures. Neutron irradiation is among the most critical factors that define the harsh service condition of LWR fuel during the normal operation. The overarching goal of the present joining and irradiation studies is to establish technologies for joining SiC-based materials for use as the LWR fuel cladding. The purpose of this work is to fabricate SiC joint specimens, characterize those joints in an unirradiated condition, and prepare rabbit capsules for neutron irradiation study on the fabricated specimens in the High Flux Isotope Reactor (HFIR). Torsional shear test specimens of chemically vapor-deposited SiC were prepared by seven different joining methods either at Oak Ridge National Laboratory or by industrial partners. The joint test specimens were characterized for shear strength and microstructures in an unirradiated condition. Rabbit irradiation capsules were designed and fabricated for neutron irradiation of these joint specimens at an LWR-relevant temperature. These rabbit capsules, already started irradiation in HFIR, are scheduled to complete irradiation to an LWR-relevant dose level in early 2015.

Silver release from coated particle fuel

International Nuclear Information System (INIS)

Brown, P.E.; Nabielek, H.

1977-03-01

The fission product Ag-110 m released from coated particles can be the dominant source of radioactivity from the core of a high temperature reactor in the early stages of the reactor life and possibly limits the accessability of primary circuit components. It can be shown that silver is retained in oxide fuel by a diffusion process (but not in carbide or carbon-diluted fuel) and that silver is released through all types of pyrocarbon layers. The retention in TRISO particles is variable and seems to be mainly connected with operating temperature and silicon carbide quality. (orig.) [de

Analysis of neutron parameters in light water moderated lattices of ThO2 and UO2 fuel rods

International Nuclear Information System (INIS)

Onusic Junior, J.; Oosterkamp, W.J.

1977-01-01

A large number of light water moderated lattices of UO 2 and ThO 2 fuel rods were analyzed with the code HAMMER. The purpose of the study was to compare experimental results with computer calculated values. The model employed is described and some modification were introduced in the resonance parameters of Th-232 to increase the agreement with the experimental value [pt

Collapse transition and cyclomatic number distribution of directed lattice animals

International Nuclear Information System (INIS)

Lam, P.M.; Duarte, J.A.M.S.

1987-01-01

The authors computed the specific heat of directed lattice animals using a Monte Carlo method for various animals sizes N, with N up to 100 on the square and N up to 125 on the simple cubic lattices. The specific heat as a function of the temperature for various animal sizes exhibits peaks which seem to approach a collapse transition temperature monotonically from below with increasing N. A least square fit together with finite-size scaling then gives both the transition temperature T/sub c/ and the specific heat exponent α for these lattices. The cyclomatic number distributions for the number of animals with fixed animal size N are also calculated and these seem to obey a scaling law for large N

Multilayer DNA Origami Packed on Hexagonal and Hybrid Lattices

OpenAIRE

Ke, Yonggang; Voigt, Niels V.; Gothelf, Kurt V.; Shih, William M.

2012-01-01

“Scaffolded DNA origami” has been proven to be a powerful and efficient approach to construct two-dimensional or three-dimensional objects with great complexity. Multilayer DNA origami has been demonstrated with helices packing along either honeycomb-lattice geometry or square-lattice geometry. Here we report successful folding of multilayer DNA origami with helices arranged on a close-packed hexagonal lattice. This arrangement yields a higher density of helical packing and therefore higher r...

In-pile tests of HTGR fuel particles and fuel elements

International Nuclear Information System (INIS)

Chernikov, A.S.; Kolesov, V.S.; Deryugin, A.I.

1985-01-01

Main types of in-pile tests for specimen tightness control at the initial step, research of fuel particle radiation stability and also study of fission product release from fuel elements during irradiation are described in this paper. Schemes and main characteristics of devices used for these tests are also given. Principal results of fission gas product release measurements satisfying HTGR demands are illustrated on the example of fuel elements, manufactured by powder metallurgy methods and having TRISO fuel particles on high temperature pyrocarbon and silicon carbide base. (author)

Lattice Boltzmann simulation for temperature-sensitive magnetic fluids in a porous square cavity

International Nuclear Information System (INIS)

Jin Licong; Zhang Xinrong; Niu Xiaodong

2012-01-01

A lattice Boltzmann method is developed to simulate temperature-sensitive magnetic fluids in a porous cavity. In the simulation, the magnetic force, efficient gravity, viscous loss term and geometric loss term in porous medium are imported to the momentum equation. To test the reliability of the method, a validation with water in porous cavity is carried out. Good agreements with the previous results verify that the present lattice Boltzmann method is promising for simulation of magnetic fluids in porous medium. In this study, we investigate the change of magnetization with external magnetic field, and we present numerical results for the streamlines, isotherms, and magnetization at vertical or horizontal mid-profiles for different values of Ram. In addition, Nusselt numbers changing with magnetic Rayleigh numbers are also investigated. - Highlights: → Developed a lattice Boltzmann method for magnetic nano-fluids in porous cavity. → Clarified flow and heat transfer for different values of (magnetic) Rayleigh numbers. → Heat transfer enhancement for magnetic fluid in porous cavity.

Fuel assembly for use in BWR type reactor

International Nuclear Information System (INIS)

Inaba, Yuzo.

1988-01-01

Purpose: To attain the reduction of neutron irradiation amount to control rods by the improvement in the reactor shutdown margin and the improvement of the control rod worth, by enhancing the arrangement of burnable poisons. Constitution: The number of burnable poison-incorporated fuel rods present in the outer two rows along the sides in adjacent with a control rod among the square lattice arrangement in a fuel assembly is decreased to less than 1/4 for that of total burnable poison-incorporated fuel rods, while the remaining burnable posion-incorporated fuel rods are arranged in the region other than above (that is, those regions not nearer to the control rod). Thus, even if a sufficient number of burnable poison to prolong the controlling effect for the reactivity with the burnable contents as the fuel assembly are disposed, only the burnable poison -incorporated fuel rods by the number less than 1/4 for that of the total burnable poison-incorporated fuel rods are present near the control rod of the fuel assembly. Accordingly, the control rod worth at the initial stage of the burning is increased at both high and normal temperatures. (Kawakami, Y.)

A report with the results of measurements carried out in AQUILON 2 for the Junta de Energia Nuclear

International Nuclear Information System (INIS)

Lourme, P.; Jacquemart, R.; Ledanois, G.

1964-01-01

Buckling measurements have been performed in support of the 'Junta de Energia Nuclear' program in the heavy water moderated critical facility AQUILON concerning lattices of uranium carbide fuel assemblies. In part of the experiments, these fuel assemblies were sheathed by housing tubes filled with organic to simulate a coolant. Each fuel assembly is a cluster of nineteen aluminium cladded, 13 mm in diameter U C rods. Experiments were made using substitution method i.e. critical approach and pulsed neutron techniques. Results are compared to lattice calculation analogous to oxide cluster calculation except about p factor. (authors) [fr

Study on niobium carbide dispersed superconducting tapes

Energy Technology Data Exchange (ETDEWEB)

Wada, H; Tachikawa, K [National Research Inst. for Metals, Tokyo (Japan); Oh' asa, M [Science Univ. of Tokyo (Japan)

1977-11-01

Niobium carbide (NbC) dispersed superconducting tapes have been fabricated by two metallurgical processes. In the first process, Ni-Nb-C alloys are directly arc melted and hot worked in air and the NbC phase is distributed in the form of fine discrete particles. In the second process, Ni-Nb and Ni-Nb-Cu alloys are arc melted, hot worked and subjected to solid-state carburization. NbC then precipitates along the grain boundaries, forming a network. The highest superconducting transition temperature attained is about 11 K. Taken together with the lattice parameter measurement, this indicates that NbC with a nearly perfect NaCl structure is formed in both processes. Measured values of the upper critical field, the critical current density and the volume fraction of the NbC phase are also discussed.

Epithermal neutron activation analysis using a boron carbide irradiation filter

International Nuclear Information System (INIS)

Ehmann, W.D.; Brueckner, J.

1980-01-01

The use of boron carbide as a thermal neutron filter in epithermal neutron activation (ENAA) analysis has been investigated. As compared to the use of a cadmium filter, boron provides a greater reduction of activities from elements relatively abundant in terrestrial rocks and fossil fuels, such as Na, La, Sc and Fe. These elements have excitation functions which follow the 1/v law in the 1 to 10 eV lower epithermal region. This enhances the sensitivity of ENAA for elements such as U, Th, Ba and etc. which have strong resonances in the higher epithermal region above 10 eV. In addition, a boron carbide filter has the advantages over cadmium of acquiring a relatively low level of induced activity which poses minimal radiation safety problems, when used for ENAA. (author)

Nuclear fuels and development of nuclear fuel elements

International Nuclear Information System (INIS)

Sundaram, C.V.; Mannan, S.L.

1989-01-01

Safe, reliable and economic operation of nuclear fission reactors, the source of nuclear power at present, requires judicious choice, careful preparation and specialised fabrication procedures for fuels and fuel element structural materials. These aspects of nuclear fuels (uranium, plutonium and their oxides and carbides), fuel element technology and structural materials (aluminium, zircaloy, stainless steel etc.) are discussed with particular reference to research and power reactors in India, e.g. the DHRUVA research reactor at BARC, Trombay, the pressurised heavy water reactors (PHWR) at Rajasthan and Kalpakkam, and the Fast Breeder Test Reactor (FBTR) at Kalpakkam. Other reactors like the gas-cooled reactors operating in UK are also mentioned. Because of the limited uranium resources, India has opted for a three-stage nuclear power programme aimed at the ultimate utilization of her abundant thorium resources. The first phase consists of natural uranium dioxide-fuelled, heavy water-moderated and cooled PHWR. The second phase was initiated with the attainment of criticality in the FBTR at Kalpakkam. Fast Breeder Reactors (FBR) utilize the plutonium and uranium by-products of phase 1. Moreover, FBR can convert thorium into fissile 233 U. They produce more fuel than is consumed - hence, the name breeders. The fuel parameters of some of the operating or proposed fast reactors in the world are compared. FBTR is unique in the choice of mixed carbides of plutonium and uranium as fuel. Factors affecting the fuel element performance and life in various reactors e.g. hydriding of zircaloys, fuel pellet-cladding interaction etc. in PHWR and void swelling; irradiation creep and helium embrittlement of fuel element structural materials in FBR are discussed along with measures to overcome some of these problems. (author). 15 refs., 9 tabs., 23 figs

Advanced Collimators for Verification of the Pu Isotopic Composition in Fresh Fuel by High Resolution Gamma Spectrometry

International Nuclear Information System (INIS)

Lebrun, Alain; Berlizov, Andriy

2013-06-01

IAEA verification of the nuclear material contained in fresh nuclear fuel assemblies is usually based on neutron coincidence counting (NCC). In the case of uranium fuel, active NCC provides the total content of uranium-235 per unit of length which, combined with active length verification, fully supports the verification. In the case of plutonium fuel, passive NCC provides the plutonium-240 equivalent content which needs to be associated with a measurement of the isotopic composition and active length measurement to complete the verification. Plutonium isotopic composition is verified by high resolution gamma spectrometry (HRGS) applied on fresh fuel assemblies assuming all fuel rods are fabricated from the same plutonium batch. For particular verifications when such an assumption cannot be reasonably made, there is a need to optimize the HRGS measurement so that contributions of internal rods to the recorded spectrum are maximized, thus providing equally strong verification of the internal fuel rods. This paper reports on simulation work carried out to design special collimators aimed at reducing the relative contribution of external fuel rods while enhancing the signal recorded from internal rods. Both cases of square lattices (e.g. 17x17 pressurized water reactor (PWR) fuel) and hexagonal compact lattices (e.g. BN800 fast neutron reactor (FNR) fuel) have been addressed. In the case of PWR lattices, the relatively large optical path to internal pins compensates for low plutonium concentrations and the large size of the fuel assemblies. A special collimator based on multiple, asymmetrical, vertical slots allows recording a spectrum from internal rods only when needed. In the FNR case, the triangular lattice is much more compact and the optical path to internal rods is very narrow. However, higher plutonium concentration and use of high energy ranges allow the verification of internal rods to be significantly strengthened. Encouraging results from the simulation

A study on the basic CVD process technology for TRISO coated particle fuel

International Nuclear Information System (INIS)

Choi, D. J.; Cheon, J. H.; Keum, I. S.; Lee, H. S.; Kim, J. G.

2006-03-01

Hydrogen energy has many advantages and is suitable as alternative energy of fossil fuel. The study of nuclear hydrogen production has performed at present. For nuclear hydrogen production, it is needed the study of VHTR(Very High Temperature Reactor) and TRISO(TRI-iSOtropic) coated fuel. TRISO coated fuel particle deposited by FBCVD(Fludized Bed CVD) method is composed of three isotropic layers: Inner Pyrolytic Carbon (IPyC), Silicon Carbide (SiC), Outer Pyrolytic Carbon (OPyC) layers. Silicon carbide was chemically vapor deposed on graphite substrate using methyltrichlorosilane (CH 3 SiCl 3 ) as a source in hydrogen atmosphere. The effect of deposition temperature and input gas ratios ( α=Q H2 /Q MTS =P H2 /P MTS ) was investigated in order to find out characteristics of silicon carbide layer. From results of those, SiC-TRISO coating deposition was conducted and achieved. Zirconium carbide layer as an advanced material of silicon carbide layer has studied. In order to find out basic properties and characteristics, studies have conducted using various methods. Zirconium carbide is chemically vapor deposed subliming zirconium tetrachloride(ZrCl 4 ) and using methan(CH 4 ) as a source in hydrogen atmosphere. Many experiments were conducted on graphite substrate about many deposition conditions such as ZrCl 4 heating temperatures and variables of H2 and CH 4 flow rate. but carbon graphite was deposited. For deposition of zirconium carbide, several different methods were approached. so zirconium carbide deposed on ZrO 2 substrate. In this experiments. source subliming type and equipment are no problems. But deposition of zirconium carbide will be continuously studied on graphite substrate approaching views of experimental way and equipment structure

The development of CVR coatings for PBR fuels

Science.gov (United States)

Barletta, R. E.; Vanier, P. E.; Dowell, M. B.; Lennartz, J. A.

Particle bed reactors (PBR's) are being developed for both space power and propulsion applications. These reactors operate with exhaust gas temperatures of 2500 to 3000 K and fuel temperatures hundreds of degrees higher. One fuel design for these reactors consists of uranium carbide encapsulated in either carbon or graphite. This fuel kernel must be protected from the coolant gas, usually H2, both to prevent attack of the kernel and to limit fission product release. Refractory carbide coatings have been proposed for this purpose. The typical coating process used for this is a chemical vapor deposition. Testing of other components have indicated the superiority of refractory carbide coatings applied using a chemical vapor reaction (CVR) process, however technology to apply these coatings to large numbers of fuel particles with diameters on the order of 500 pm were not readily available. A process to deposit these CVR coatings on surrogate fuel consisting of graphite particles is described. Several types of coatings have been applied to the graphite substrate: NbC in various thicknesses and a bilayer coating consisting of NbC and TaC with a intermediate layer of pyrolytic graphite. These coated particles have been characterized prior to test; results are presented.

Comparison of heuristic optimization techniques for the enrichment and gadolinia distribution in BWR fuel lattices and decision analysis

International Nuclear Information System (INIS)

Castillo, Alejandro; Martín-del-Campo, Cecilia; Montes-Tadeo, José-Luis; François, Juan-Luis; Ortiz-Servin, Juan-José; Perusquía-del-Cueto, Raúl

2014-01-01

Highlights: • Different metaheuristic optimization techniques were compared. • The optimal enrichment and gadolinia distribution in a BWR fuel lattice was studied. • A decision making tool based on the Position Vector of Minimum Regret was applied. • Similar results were found for the different optimization techniques. - Abstract: In the present study a comparison of the performance of five heuristic techniques for optimization of combinatorial problems is shown. The techniques are: Ant Colony System, Artificial Neural Networks, Genetic Algorithms, Greedy Search and a hybrid of Path Relinking and Scatter Search. They were applied to obtain an “optimal” enrichment and gadolinia distribution in a fuel lattice of a boiling water reactor. All techniques used the same objective function for qualifying the different distributions created during the optimization process as well as the same initial conditions and restrictions. The parameters included in the objective function are the k-infinite multiplication factor, the maximum local power peaking factor, the average enrichment and the average gadolinia concentration of the lattice. The CASMO-4 code was used to obtain the neutronic parameters. The criteria for qualifying the optimization techniques include also the evaluation of the best lattice with burnup and the number of evaluations of the objective function needed to obtain the best solution. In conclusion all techniques obtain similar results, but there are methods that found better solutions faster than others. A decision analysis tool based on the Position Vector of Minimum Regret was applied to aggregate the criteria in order to rank the solutions according to three functions: neutronic grade at 0 burnup, neutronic grade with burnup and global cost which aggregates the computing time in the decision. According to the results Greedy Search found the best lattice in terms of the neutronic grade at 0 burnup and also with burnup. However, Greedy Search is

Nuclear reactor fuel element

International Nuclear Information System (INIS)

D'Eye, R.W.M.; Shennan, J.V.; Ford, L.H.

1977-01-01

Fuel element with particles from ceramic fissionable material (e.g. uranium carbide), each one being coated with pyrolitically deposited carbon and all of them being connected at their points of contact by means of an individual crossbar. The crossbar consists of silicon carbide produced by reaction of silicon metal powder with the carbon under the influence of heat. Previously the silicon metal powder together with the particles was kneaded in a solvent and a binder (e.g. epoxy resin in methyl ethyl ketone plus setting agent) to from a pulp. The reaction temperature lies at 1750 0 C. The reaction itself may take place in a nitrogen atmosphere. There will be produced a fuel element with a high overall thermal conductivity. (DG) [de

Thermal-hydraulic performance analysis of a subchannel with square and triangle fuel rod arrangements using the entropy generation approach

Institute of Scientific and Technical Information of China (English)

S.Talebi; M.M.Valoujerdi

2017-01-01

The present paper discusses entropy generation in fully developed turbulent flows through a subchannel,arranged in square and triangle arrays.Entropy generation is due to contribution of both heat transfer and pressure drop.Our main objective is to study the effect of key parameters such as spacer grid,fuel rod power distribution,Reynolds number Re,dimensionless heat power ω,lengthto-fuel-diameter ratio λ,and pitch-to-diameter ratio ξ on subchannel entropy generation.The analysis explicitly shows the contribution of heat transfer and pressure drop to the total entropy generation.An analytical formulation is introduced to total entropy generation for situations with uniform and sinusoidal rod power distribution.It is concluded that power distribution affects entropy generation.A smoother power profile leads to less entropy generation.The entropy generation of square rod array bundles is more efficient than that of triangular rod arrays,and spacer grids generate more entropy.

Quantum phases of dipolar rotors on two-dimensional lattices.

Science.gov (United States)

Abolins, B P; Zillich, R E; Whaley, K B

2018-03-14

The quantum phase transitions of dipoles confined to the vertices of two-dimensional lattices of square and triangular geometry is studied using path integral ground state quantum Monte Carlo. We analyze the phase diagram as a function of the strength of both the dipolar interaction and a transverse electric field. The study reveals the existence of a class of orientational phases of quantum dipolar rotors whose properties are determined by the ratios between the strength of the anisotropic dipole-dipole interaction, the strength of the applied transverse field, and the rotational constant. For the triangular lattice, the generic orientationally disordered phase found at zero and weak values of both dipolar interaction strength and applied field is found to show a transition to a phase characterized by net polarization in the lattice plane as the strength of the dipole-dipole interaction is increased, independent of the strength of the applied transverse field, in addition to the expected transition to a transverse polarized phase as the electric field strength increases. The square lattice is also found to exhibit a transition from a disordered phase to an ordered phase as the dipole-dipole interaction strength is increased, as well as the expected transition to a transverse polarized phase as the electric field strength increases. In contrast to the situation with a triangular lattice, on square lattices, the ordered phase at high dipole-dipole interaction strength possesses a striped ordering. The properties of these quantum dipolar rotor phases are dominated by the anisotropy of the interaction and provide useful models for developing quantum phases beyond the well-known paradigms of spin Hamiltonian models, implementing in particular a novel physical realization of a quantum rotor-like Hamiltonian that possesses an anisotropic long range interaction.

Quantum phases of dipolar rotors on two-dimensional lattices

Science.gov (United States)

Abolins, B. P.; Zillich, R. E.; Whaley, K. B.

2018-03-01

The quantum phase transitions of dipoles confined to the vertices of two-dimensional lattices of square and triangular geometry is studied using path integral ground state quantum Monte Carlo. We analyze the phase diagram as a function of the strength of both the dipolar interaction and a transverse electric field. The study reveals the existence of a class of orientational phases of quantum dipolar rotors whose properties are determined by the ratios between the strength of the anisotropic dipole-dipole interaction, the strength of the applied transverse field, and the rotational constant. For the triangular lattice, the generic orientationally disordered phase found at zero and weak values of both dipolar interaction strength and applied field is found to show a transition to a phase characterized by net polarization in the lattice plane as the strength of the dipole-dipole interaction is increased, independent of the strength of the applied transverse field, in addition to the expected transition to a transverse polarized phase as the electric field strength increases. The square lattice is also found to exhibit a transition from a disordered phase to an ordered phase as the dipole-dipole interaction strength is increased, as well as the expected transition to a transverse polarized phase as the electric field strength increases. In contrast to the situation with a triangular lattice, on square lattices, the ordered phase at high dipole-dipole interaction strength possesses a striped ordering. The properties of these quantum dipolar rotor phases are dominated by the anisotropy of the interaction and provide useful models for developing quantum phases beyond the well-known paradigms of spin Hamiltonian models, implementing in particular a novel physical realization of a quantum rotor-like Hamiltonian that possesses an anisotropic long range interaction.

Physicochemical analysis of interaction of oxide fuel with pyrocarbon coatings of fuel particles

International Nuclear Information System (INIS)

Lyutikov, R.A.; Khromov, Yu.F.; Chernikov, A.S.

1990-01-01

Equilibrium pressure of (CO+Kr,Xe) gases inside fuel particle with oxide kern depending on design features of fuel particle, on temperature. on (O/U) initial composition and fuel burnup is calculated using the suggested model. Analysis of possibility for gas pressure reduction by means of uranium carbide alloying of kern and degree increase of solid fission product retention (Cs for example) during alumosilicate alloying of uranium oxide is conducted

Formation mechanism of dot-line square superlattice pattern in dielectric barrier discharge

Energy Technology Data Exchange (ETDEWEB)

Liu, Weibo; Dong, Lifang, E-mail: donglfhbu@163.com, E-mail: pyy1616@163.com; Wang, Yongjie; Zhang, Xinpu [College of Physics Science and Technology, Hebei University, Baoding 071002 (China); College of Quality and Technical Supervision, Hebei University, Baoding 071002 (China); Pan, Yuyang, E-mail: donglfhbu@163.com, E-mail: pyy1616@163.com [College of Quality and Technical Supervision, Hebei University, Baoding 071002 (China)

2014-11-15

We investigate the formation mechanism of the dot-line square superlattice pattern (DLSSP) in dielectric barrier discharge. The spatio-temporal structure studied by using the intensified-charge coupled device camera shows that the DLSSP is an interleaving of three different subpatterns in one half voltage cycle. The dot square lattice discharges first and, then, the two kinds of line square lattices, which form square grid structures discharge twice. When the gas pressure is varied, DLSSP can transform from square superlattice pattern (SSP). The spectral line profile method is used to compare the electron densities, which represent the amounts of surface charges qualitatively. It is found that the amount of surface charges accumulated by the first discharge of DLSSP is less than that of SSP, leading to a bigger discharge area of the following discharge (lines of DLSSP instead of halos of SSP). The spatial distribution of the electric field of the surface charges is simulated to explain the formation of DLSSP. This paper may provide a deeper understanding for the formation mechanism of complex superlattice patterns in DBD.

REFEL silicon carbide. The development of a ceramic for a nuclear engineering application

Energy Technology Data Exchange (ETDEWEB)

Kennedy, P.; Shennan, J. V.

1974-10-15

REFEL silicon carbide is a strong, uniform, fine-grain material which retains its strength and is stable in an oxidizing environment up to 1400 deg C. REFEL silicon carbide tube can be produced in quantity and by a combination of process controls, visual examination, NDT and proof testing, a very consistent product can be made. The material was developed as a nuclear fuel cladding capable of operating at temperatures o 1100 deg C in a CO₂-cooled reactor and the combination of excellent physical, mechanical and chemical properties together with product consistency ave confirmed the feasibility of this application. In a series of irradiation experiments, REFEL silicon carbide clad fuel pins have behaved predictably. At irradiation temperatures below about 800 deg C, the thermal conductivity falls sharply, the associate thermal stress increases, and the probability of failure, for the same rating, increases. It has been demonstrated theoretically that this effect can be overcome by halving the tube wall thickness. In addition to the thermal stress enhancement, the strength and Weibull modulus also fall under irradiation and consequently the safe working stress is reduced, Calculations show that in the absence of irradiation a fourfold increase in rating cold be tolerated. Thus, the material should have excellent thermal stress resistance in non-nuclear applications such as gas turbine components. (auth)

Porous silicon carbide (SIC) semiconductor device

Science.gov (United States)

Shor, Joseph S. (Inventor); Kurtz, Anthony D. (Inventor)

1996-01-01

Porous silicon carbide is fabricated according to techniques which result in a significant portion of nanocrystallites within the material in a sub 10 nanometer regime. There is described techniques for passivating porous silicon carbide which result in the fabrication of optoelectronic devices which exhibit brighter blue luminescence and exhibit improved qualities. Based on certain of the techniques described porous silicon carbide is used as a sacrificial layer for the patterning of silicon carbide. Porous silicon carbide is then removed from the bulk substrate by oxidation and other methods. The techniques described employ a two-step process which is used to pattern bulk silicon carbide where selected areas of the wafer are then made porous and then the porous layer is subsequently removed. The process to form porous silicon carbide exhibits dopant selectivity and a two-step etching procedure is implemented for silicon carbide multilayers.

Chirality in distorted square planar Pd(O,N)2 compounds.

Science.gov (United States)

Brunner, Henri; Bodensteiner, Michael; Tsuno, Takashi

2013-10-01

Salicylidenimine palladium(II) complexes trans-Pd(O,N)2 adopt step and bowl arrangements. A stereochemical analysis subdivides 52 compounds into 41 step and 11 bowl types. Step complexes with chiral N-substituents and all the bowl complexes induce chiral distortions in the square planar system, resulting in Δ/Λ configuration of the Pd(O,N)2 unit. In complexes with enantiomerically pure N-substituents ligand chirality entails a specific square chirality and only one diastereomer assembles in the lattice. Dimeric Pd(O,N)2 complexes with bridging N-substituents in trans-arrangement are inherently chiral. For dimers different chirality patterns for the Pd(O,N)2 square are observed. The crystals contain racemates of enantiomers. In complex two independent molecules form a tight pair. The (RC) configuration of the ligand induces the same Δ chirality in the Pd(O,N)2 units of both molecules with varying square chirality due to the different crystallographic location of the independent molecules. In complexes and atrop isomerism induces specific configurations in the Pd(O,N)2 bowl systems. The square chirality is largest for complex [(Diop)Rh(PPh3 )Cl)], a catalyst for enantioselective hydrogenation. In the lattice of two diastereomers with the same (RC ,RC) configuration in the ligand Diop but opposite Δ and Λ square configurations co-crystallize, a rare phenomenon in stereochemistry. © 2013 Wiley Periodicals, Inc.

Inflection points of microcanonical entropy: Monte Carlo simulation of q state Potts model on a finite square lattice

Energy Technology Data Exchange (ETDEWEB)

Praveen, E., E-mail: svmstaya@gmail.com; Satyanarayana, S. V. M., E-mail: svmstaya@gmail.com [Department of Physics, Pondicherry University, Puducherry-605014 (India)

2014-04-24

Traditional definition of phase transition involves an infinitely large system in thermodynamic limit. Finite systems such as biological proteins exhibit cooperative behavior similar to phase transitions. We employ recently discovered analysis of inflection points of microcanonical entropy to estimate the transition temperature of the phase transition in q state Potts model on a finite two dimensional square lattice for q=3 (second order) and q=8 (first order). The difference of energy density of states (DOS) Δ ln g(E) = ln g(E+ ΔE) −ln g(E) exhibits a point of inflexion at a value corresponding to inverse transition temperature. This feature is common to systems exhibiting both first as well as second order transitions. While the difference of DOS registers a monotonic variation around the point of inflexion for systems exhibiting second order transition, it has an S-shape with a minimum and maximum around the point of inflexion for the case of first order transition.

Inflection points of microcanonical entropy: Monte Carlo simulation of q state Potts model on a finite square lattice

International Nuclear Information System (INIS)

Praveen, E.; Satyanarayana, S. V. M.

2014-01-01

Traditional definition of phase transition involves an infinitely large system in thermodynamic limit. Finite systems such as biological proteins exhibit cooperative behavior similar to phase transitions. We employ recently discovered analysis of inflection points of microcanonical entropy to estimate the transition temperature of the phase transition in q state Potts model on a finite two dimensional square lattice for q=3 (second order) and q=8 (first order). The difference of energy density of states (DOS) Δ ln g(E) = ln g(E+ ΔE) −ln g(E) exhibits a point of inflexion at a value corresponding to inverse transition temperature. This feature is common to systems exhibiting both first as well as second order transitions. While the difference of DOS registers a monotonic variation around the point of inflexion for systems exhibiting second order transition, it has an S-shape with a minimum and maximum around the point of inflexion for the case of first order transition

Fluidized bed deposition and evaluation of silicon carbide coatings on microspheres

International Nuclear Information System (INIS)

Federer, J.I.

1977-01-01

The fuel element for the HTGR is an array of closely packed fuel microspheres in a carbonaceous matrix. A coating of dense silicon carbide (SiC), along with pyrocarbon layers, is deposited on the fueled microspheres to serve as a barrier against diffusion of fission products. The microspheres are coated with silicon carbide in a fluidized bed by reaction of methyltrichlorosilane (CH 3 SiCl 3 or MTS) and hydrogen at elevated temperatures. The principal variables of coating temperature and reactant gas composition (H 2 /MTS ratio) have been correlated with coating rate, morphology, stoichiometry, microstructure, and density. The optimum temperature for depositing highly dense coatings is in the range 1475 to 1675 0 C. Lower temperatures result in silicon-rich deposits, while higher temperatures may cause unacceptable porosity. The optimum H 2 /MTS ratio for highly dense coatings is 20 or more (approximately 5% MTS or less). The amount of grown-in porosity increases as the H 2 /MTS ratio decreases below 20. The requirement that the H 2 /MTS ratio be about 20 or more imposes a practical restraint on coating rate, since increasing the total flow rate would eventually expel microspheres from the coating tube. Evaluation of stoichiometry, morphology, and microstructure support the above mentioned optimum conditions of temperature and reactant gas composition. 18 figures, 3 tables

Enhancing the brightness of electrically driven single-photon sources using color centers in silicon carbide

Science.gov (United States)

Khramtsov, Igor A.; Vyshnevyy, Andrey A.; Fedyanin, Dmitry Yu.

2018-03-01

Practical applications of quantum information technologies exploiting the quantum nature of light require efficient and bright true single-photon sources which operate under ambient conditions. Currently, point defects in the crystal lattice of diamond known as color centers have taken the lead in the race for the most promising quantum system for practical non-classical light sources. This work is focused on a different quantum optoelectronic material, namely a color center in silicon carbide, and reveals the physics behind the process of single-photon emission from color centers in SiC under electrical pumping. We show that color centers in silicon carbide can be far superior to any other quantum light emitter under electrical control at room temperature. Using a comprehensive theoretical approach and rigorous numerical simulations, we demonstrate that at room temperature, the photon emission rate from a p-i-n silicon carbide single-photon emitting diode can exceed 5 Gcounts/s, which is higher than what can be achieved with electrically driven color centers in diamond or epitaxial quantum dots. These findings lay the foundation for the development of practical photonic quantum devices which can be produced in a well-developed CMOS compatible process flow.

Numerical investigation of heat transfer in upward flows of supercritical water in circular tubes and tight fuel rod bundles

International Nuclear Information System (INIS)

Yang Jue; Oka, Yoshiaki; Ishiwatari, Yuki; Liu Jie; Yoo, Jaewoon

2007-01-01

Heat transfer in upward flows of supercritical water in circular tubes and in tight fuel rod bundles is numerically investigated by using the commercial CFD code STAR-CD 3.24. The objective is to have more understandings about the phenomena happening in supercritical water and for designs of supercritical water cooled reactors. Some turbulence models are selected to carry out numerical simulations and the results are compared with experimental data and other correlations to find suitable models to predict heat transfer in supercritical water. The comparisons are not only in the low bulk temperature region, but also in the high bulk temperature region. The two-layer model (Hassid and Poreh) gives a better prediction to the heat transfer than other models, and the standard k-ε high Re model with the standard wall function also shows an acceptable predicting capability. Three-dimensional simulations are carried out in sub-channels of tight square lattice and triangular lattice fuel rod bundles at supercritical pressure. Results show that there is a strong non-uniformity of the circumferential distribution of the cladding surface temperature, in the square lattice bundle with a small pitch-to-diameter ratio (P/D). However, it does not occur in the triangular lattice bundle with a small P/D. It is found that this phenomenon is caused by the large non-uniformity of the flow area in the cross-section of sub-channels. Some improved designs are numerically studied and proved to be effective to avoid the large circumferential temperature gradient at the cladding surface

Review of thermal expansion and density of uranium and plutonium carbides

International Nuclear Information System (INIS)

Andrew, J.F.; Latimer, T.W.

1975-07-01

The published literature on linear thermal expansion and density of uranium and plutonium carbide nuclear fuels, including UC, PuC, (U,Pu)C, U 2 C 3 , Pu 2 C 3 , and (U,Pu) 2 C 3 , is critically reviewed. Recommended values are given in tabular form and additional experimental studies needed for completeness are outlined. 16 tables, 52 references

Fuel cycles with high fuel burn-up: analysis of reactivity coefficients

Energy Technology Data Exchange (ETDEWEB)

Kryuchkov, E.F.; Shmelev, A.N.; Ternovykh, M.J.; Tikhomirov, G.V.; Jinhong, L. [Moscow Engineering Physics Institute (State University) (Russian Federation); Saito, M. [Tokyo Institute of Technology (Japan)

2003-07-01

Fuel cycles of light-water reactors (LWR) with high fuel burn-up (above 100 MWd/kg), as a rule, involve large amounts of fissionable materials. It leads to forming the neutron spectrum harder than that in traditional LWR. Change of neutron spectrum and significant amount of non-traditional isotopes (for example, {sup 237}Np, {sup 238}Pu, {sup 231}Pa, {sup 232}U) in such fuel compositions can alter substantially reactivity coefficients as compared with traditional uranium-based fuel. The present work addresses the fuel cycles with high fuel burn-up which are based on Th-Pa-U and U-Np-Pu fuel compositions. Numerical analyses are carried out to determine effective neutron multiplication factor and void reactivity coefficient (VRC) for different values of fuel burn-up and different lattice parameters. The algorithm is proposed for analysis of isotopes contribution to these coefficients. Various ways are considered to upgrade safety of nuclear fuel cycles with high fuel burn-up. So, the results obtained in this study have demonstrated that: -1) Non-traditional fuel compositions developed for achievement of high fuel burn-up in LWR can possess positive values of reactivity coefficients that is unacceptable from the reactor operation safety point of view; -2) The lattice pitch of traditional LWR is not optimal for non-traditional fuel compositions, the increased value of the lattice pitch leads to larger value of initial reactivity margin and provides negative VRC within sufficiently broad range of coolant density; -3) Fuel burn-up has an insignificant effect on VRC dependence on coolant density, so, the measures undertaken to suppress positive VRC of fresh fuel will be effective for partially burnt-up fuel compositions also and; -4) Increase of LWR core height and introduction of additional moderators into the fuel lattice can be used as the ways to reach negative VRC values for full range of possible coolant density variations.

Calculational methods for lattice cells

International Nuclear Information System (INIS)

Askew, J.R.

1980-01-01

At the current stage of development, direct simulation of all the processes involved in the reactor to the degree of accuracy required is not an economic proposition, and this is achieved by progressive synthesis of models for parts of the full space/angle/energy neutron behaviour. The split between reactor and lattice calculations is one such simplification. Most reactors are constructed of repetitions of similar geometric units, the fuel elements, having broadly similar properties. Thus the provision of detailed predictions of their behaviour is an important step towards overall modelling. We shall be dealing with these lattice methods in this series of lectures, but will refer back from time to time to their relationship with overall reactor calculation The lattice cell is itself composed of somewhat similar sub-units, the fuel pins, and will itself often rely upon a further break down of modelling. Construction of a good model depends upon the identification, on physical and mathematical grounds, of the most helpful division of the calculation at this level

Plasma spraying of zirconium carbide – hafnium carbide – tungsten cermets

Czech Academy of Sciences Publication Activity Database

Brožek, Vlastimil; Ctibor, Pavel; Cheong, D.-I.; Yang, S.-H.

2009-01-01

Roč. 9, č. 1 (2009), s. 49-64 ISSN 1335-8987 Institutional research plan: CEZ:AV0Z20430508 Keywords : Plasma spraying * cermet coatings * microhardness * zirconium carbide * hafnium carbide * tungsten * water stabilized plasma Subject RIV: JH - Ceramics, Fire-Resistant Materials and Glass

Two dimentional lattice vibrations from direct product representations of symmetry groups

Directory of Open Access Journals (Sweden)

J. N. Boyd

1983-01-01

two dimensional crystals. First, the Born cyclic condition is applied to a double chain composed of coupled linear lattices to obtain a cylindrical arrangement. Then the quadratic Lagrangian function for the system is written in matrix notation. The Lagrangian is diagonalized to yield the natural frequencies of the system. The transformation to achieve the diagonalization was obtained from group theorectic considerations. Next, the techniques developed for the double chain are applied to a square lattice. The square lattice is transformed into the toroidal Ising model. The direct product nature of the symmetry group of the torus reveals the transformation to diagonalize the Lagrangian for the Ising model, and the natural frequencies for the principal directions in the model are obtained in closed form.

Metal Carbides for Biomass Valorization

Directory of Open Access Journals (Sweden)

Carine E. Chan-Thaw

2018-02-01

Full Text Available Transition metal carbides have been utilized as an alternative catalyst to expensive noble metals for the conversion of biomass. Tungsten and molybdenum carbides have been shown to be effective catalysts for hydrogenation, hydrodeoxygenation and isomerization reactions. The satisfactory activities of these metal carbides and their low costs, compared with noble metals, make them appealing alternatives and worthy of further investigation. In this review, we succinctly describe common synthesis techniques, including temperature-programmed reaction and carbothermal hydrogen reduction, utilized to prepare metal carbides used for biomass transformation. Attention will be focused, successively, on the application of transition metal carbide catalysts in the transformation of first-generation (oils and second-generation (lignocellulose biomass to biofuels and fine chemicals.

Superconductivity in the Penson-Kolb Model on a Triangular Lattice

Science.gov (United States)

Ptok, A.; Mierzejewski, M.

2008-07-01

We investigate properties of the two-dimensional Penson-Kolb model with repulsive pair hopping interaction. In the case of a bipartite square lattice this interaction may lead to the η-type pairing, when the phase of superconducting order parameter changes from one lattice site to the neighboring one. We show that this interaction may be responsible for the onset of superconductivity also for a triangular lattice. We discuss the spatial dependence of the superconducting order parameter and demonstrate that the total momentum of the paired electrons is determined by the lattice geometry.

BCS @ 50: derivation of gap equations in different lattice geometries

International Nuclear Information System (INIS)

Saurabh Basu

2007-07-01

We rigorously derive BCS gap equations for a square, triangular and a honeycomb lattice using a two-dimensional t-J model. The gap equations in all the three lattice geometries look usual, with band indices appearing and a minor modification in the separable pair potential for the (two band) honeycomb lattice. In each case, the gap equation is solved (self consistently with the number equation) at low densities assuming singlet pairing. (author)

NOVEL SUPPORTED BIMETALLIC CARBIDE CATALYSTS FOR COPROCESSING OF COAL WITH WASTE METERIALS

Energy Technology Data Exchange (ETDEWEB)

S. Ted Oyama; David F. Cox; Chunshan Song; Fred Allen; Weilin Wang; Viviane Schwartz; Xinqin Wang; Jianli Yang

2001-01-01

The overall objectives of this project are to explore the potential of novel monometallic and bimetallic Mo-based carbide catalysts for heavy hydrocarbon coprocessing, and to understand the fundamental chemistry related to the reaction pathways of coprocessing and the role of the catalysts in the conversion of heavy hydrocarbon resources into liquid fuels based on the model compound reactions.

Alternative fuels for the French fast breeder reactors programme

International Nuclear Information System (INIS)

Bailly, H.; Bernard, H.; Mansard, B.

1989-01-01

French fast breeder reactors use mixed oxide as reference fuel. A great deal of experience has been gained in the behaviour and manufacture of oxide fuel, which has proved to be the most suitable fuel for future commercial breeder reactors. However, France is maintaining long-term alternative fuels programme, in order to be in a position to satisfy eventually new future reactor design and operational requirements. Initially, the CEA in France developed a carbide-based, sodium-bonded fuel designed for a high specific power. The new objective of the alternative fuels programme is to define a fuel which could replace the oxide without requiring any significant changes to the operating conditions, fuel cycle processes or facilities. The current program concentrates on a nitride-based, helium-bonded fuel, bearing in mind the carbide solution. The paper describes the main characteristics required, the manufacturing process as developed, the inspection methods, and the results obtained. Present indications are that the industrial manufacture of mixed nitride is feasible and that production costs for nitride and oxide fuels would be not significantly different. (author) 8 refs., 2 figs

Pattern formation in two-dimensional square-shoulder systems

International Nuclear Information System (INIS)

Fornleitner, Julia; Kahl, Gerhard

2010-01-01

Using a highly efficient and reliable optimization tool that is based on ideas of genetic algorithms, we have systematically studied the pattern formation of the two-dimensional square-shoulder system. An overwhelming wealth of complex ordered equilibrium structures emerge from this investigation as we vary the shoulder width. With increasing pressure three structural archetypes could be identified: cluster lattices, where clusters of particles occupy the sites of distorted hexagonal lattices, lane formation, and compact particle arrangements with high coordination numbers. The internal complexity of these structures increases with increasing shoulder width.

Pattern formation in two-dimensional square-shoulder systems

Energy Technology Data Exchange (ETDEWEB)

Fornleitner, Julia [Institut fuer Festkoerperforschung, Forschungsszentrum Juelich, D-52425 Juelich (Germany); Kahl, Gerhard, E-mail: fornleitner@cmt.tuwien.ac.a [Institut fuer Theoretische Physik and Centre for Computational Materials Science (CMS), Technische Universitaet Wien, Wiedner Hauptstrasse 8-10, A-1040 Wien (Austria)

2010-03-17

Using a highly efficient and reliable optimization tool that is based on ideas of genetic algorithms, we have systematically studied the pattern formation of the two-dimensional square-shoulder system. An overwhelming wealth of complex ordered equilibrium structures emerge from this investigation as we vary the shoulder width. With increasing pressure three structural archetypes could be identified: cluster lattices, where clusters of particles occupy the sites of distorted hexagonal lattices, lane formation, and compact particle arrangements with high coordination numbers. The internal complexity of these structures increases with increasing shoulder width.

Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

Energy Technology Data Exchange (ETDEWEB)

Armstrong, J.; Hamilton, H.; Hyland, B. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada)

2013-07-01

A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

International Nuclear Information System (INIS)

Armstrong, J.; Hamilton, H.; Hyland, B.

2013-01-01

A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

High reliability fuel in the US

International Nuclear Information System (INIS)

Neuhold, R.J.; Leggett, R.D.; Walters, L.C.; Matthews, R.B.

1986-05-01

The fuels development program of the United States is described for liquid metal reactors (LMR's). The experience base, status and future potential are discussed for the three systems - oxide, metal and carbide - that have proved to have high reliability. Information is presented showing burnup capability of the oxide fuel system in a large core, e.g., FFTF, to be 150 MWd/kgM with today's technology with the potential for a capability as high as 300 MWd/kgM. Data provided for the metal fuel system show 8 at. % being routinely achieved as the EBR-II driver fuel with good potential for extending this to 15 at. % since special test pins have already exceeded this burnup level. The data included for the carbide fuel system are from pin and assembly irradiations in EBR-II and FFTF, respectively. Burnup to 12 at. % appears readily achievable with burnups to 20 at. % being demonstrated in a few pins. Efforts continue on all three systems with the bulk of the activity on metal and oxide

Computer code TOBUNRAD for PWR fuel bundle heat-up calculations

International Nuclear Information System (INIS)

Shimooke, Takanori; Yoshida, Kazuo

1979-05-01

The computer code TOBUNRAD developed is for analysis of ''fuel-bundle'' heat-up phenomena in a loss-of-coolant accident of PWR. The fuel bundle consists of fuel pins in square lattice; its behavior is different from that of individual pins during heat-up. The code is based on the existing TOODEE2 code which analyzes heat-up phenomena of single fuel pins, so that the basic models of heat conduction and transfer and coolant flow are the same as the TOODEE2's. In addition to the TOODEE2 features, unheated rods are modeled and radiation heat loss is considered between fuel pins, a fuel pin and other heat sinks. The TOBUNRAD code is developed by a new FORTRAN technique which makes it possible to interrupt a flow of program controls wherever desired, thereby attaching several subprograms to the main code. Users' manual for TOBUNRAD is presented: The basic program-structure by interruption method, physical and computational model in each sub-code, usage of the code and sample problems. (author)

Comparison of MCNP and WIMS-AECL/RFSP calculations against critical heavy water experiments in ZED-2 with CANFLEX-LVRF and CANFLEX-LEU fuels

International Nuclear Information System (INIS)

Bromley, B. P.; Watts, D. G.; Pencer, J.; Zeller, M.; Dweiri, Y.

2009-01-01

This paper summarizes calculations of MCNP5 and WIMS-AECL/RFSP compared against measurements in coolant void substitution experiments in the ZED-2 critical facility with CANFLEX R-LEU/RU (Low Enriched Uranium, Recovered Uranium) reference fuels and CANFLEX-LVRF (Low Void Reactivity Fuel) test fuel, and H 2 O/air coolants. Both codes are tested for the prediction of the change in reactivity with complete voiding of all fuel channels, and that for a checkerboard voiding pattern. Understanding these phenomena is important for the ACR-1000 R reactor. Comparisons are also made for the prediction of the axial and radial neutron flux distributions, as measured by copper foil activation. The experimental data for these comparisons were obtained from critical mixed lattice / substitution experiments in AECL's ZED-2 critical facility using CANFLEX-LEU/RU and CANFLEX-LVRF fuel in a 24-cm square lattice pitch at 25 degrees C. Substitution analyses were performed to isolate the properties (buckling, bare critical lattice dimensions) of the CANFLEX-LVRF fuel. This data was then used to further test the lattice physics codes. These comparisons establish biases/uncertainties and errors in the calculation of k eff , coolant void reactivity, checkerboard coolant void reactivity, and flux distributions. Results show small to modest biases in void reactivity and very good agreement for flux distributions. The importance of boundary conditions and the modeling of un-moderated fuel in the critical experiments are demonstrated. This comparison study provides data that supports code validation and gives good confidence in the reactor physics tools used in the design and safety analysis of the ACR-1000 reactor. (authors)

Advanced Measurements of Silicon Carbide Ceramic Matrix Composites

Energy Technology Data Exchange (ETDEWEB)

Farhad Farzbod; Stephen J. Reese; Zilong Hua; Marat Khafizov; David H. Hurley

2012-08-01

Silicon carbide (SiC) is being considered as a fuel cladding material for accident tolerant fuel under the Light Water Reactor Sustainability (LWRS) Program sponsored by the Nuclear Energy Division of the Department of Energy. Silicon carbide has many potential advantages over traditional zirconium based cladding systems. These include high melting point, low susceptibility to corrosion, and low degradation of mechanical properties under neutron irradiation. In addition, ceramic matrix composites (CMCs) made from SiC have high mechanical toughness enabling these materials to withstand thermal and mechanical shock loading. However, many of the fundamental mechanical and thermal properties of SiC CMCs depend strongly on the fabrication process. As a result, extrapolating current materials science databases for these materials to nuclear applications is not possible. The “Advanced Measurements” work package under the LWRS fuels pathway is tasked with the development of measurement techniques that can characterize fundamental thermal and mechanical properties of SiC CMCs. An emphasis is being placed on development of characterization tools that can used for examination of fresh as well as irradiated samples. The work discuss in this report can be divided into two broad categories. The first involves the development of laser ultrasonic techniques to measure the elastic and yield properties and the second involves the development of laser-based techniques to measurement thermal transport properties. Emphasis has been placed on understanding the anisotropic and heterogeneous nature of SiC CMCs in regards to thermal and mechanical properties. The material properties characterized within this work package will be used as validation of advanced materials physics models of SiC CMCs developed under the LWRS fuels pathway. In addition, it is envisioned that similar measurement techniques can be used to provide process control and quality assurance as well as measurement of

Fuel performance of DOE fuels in water storage

International Nuclear Information System (INIS)

Hoskins, A.P.; Scott, J.G.; Shelton-Davis, C.V.; McDannel, G.E.

1993-01-01

Westinghouse Idaho Nuclear Company operates the Idaho Chemical Processing Plant (ICPP) at the Idaho National Engineering Laboratory. In April of 1992, the U.S. Department of Energy (DOE) decided to end the fuel reprocessing mission at ICPP. Fuel performance in storage received increased emphasis as the fuel now needs to be stored until final dispositioning is defined and implemented. Fuels are stored in four main areas: an original underwater storage facility, a modern underwater storage facility, and two dry fuel storage facilities. As a result of the reactor research mission of the DOE and predecessor agencies, the Energy Research and Development Administration and the Atomic Energy Commission, many types of nuclear fuel have been developed, used, and assigned to storage at the ICPP. Fuel clad with stainless steel, zirconium, aluminum, and graphite are represented. Fuel matrices include uranium oxide, hydride, carbide, metal, and alloy fuels, resulting in 55 different fuel types in storage. Also included in the fuel storage inventory is canned scrap material

Behaviour of a VVER-1000 fuel element with boron carbide/steel absorber tested under severe fuel damage conditions in the CORA facility (Results of experiment CORA-W2)

International Nuclear Information System (INIS)

Hagen, S.; Hofmann, P.; Noack, V.; Schanz, G.; Schumacher, G.; Sepold, L.

1994-10-01

The 'Severe Fuel Damage' (SFD) experiments of the Kernforschungszentrum Karlsruhe (KfK), Federal Republic of Germany, were carried out in the out-of-pile facility 'CORA' as part of the international Severe Fuel Damage (SFD) research. The experimental program was set up to provide information on the failure mechanisms of Light Water Reactor (LWR) fuel elements in a temperature range from 1200 C to 2000 C and in few cases up to 2400 C. Between 1987 and 1992 a total of 17 CORA experiments with two different bundle configurations, i.e. PWR (Pressurized Water Reactor) and BWR (Boiling Water Reactor) bundles were performed. These assemblies represented 'Western-type' fuel elements with the pertinent materials for fuel, cladding, grid spacer, and absorber rod. At the end of the experimental program two VVER-1000 specific tests were run in the CORA facility with identical objectives but with genuine VVER-type materials. The experiments, designated CORA-W1 and CORA-W2 were conducted on February 18, 1993 and April 21, 1993, respectively. Test bundle CORA-W1 was without absorber material whereas CORA-W2 contained one absorber rod (boron carbide/steel). As in the earlier CORA tests the test bundles were subjected to temperature transients of a slow heatup rate in a steam environment. The transient phases of the tests were initiated with a temperature ramp rate of 1 K/s. With these conditions a so-called small-break LOCA was simulated. The temperature escalation due to the exothermal zircon/niobium-steam reaction started at about 1200 C, leading the bundles to maximum temperatures of approximately 1900 C. The thermal response of bundle CORA-W2 is comparable to that of CORA-W1. In test CORA-W2, however, the temperature front moved faster from the top to the bottom compared to test CORA-W1 [de

The Formation of Carbide-Free Bainite in High-Carbon High-Silicon Steel under Isothermal Conditions

Science.gov (United States)

Tereshchenko, N. A.; Yakovleva, I. L.; Mirzaev, D. A.; Buldashev, I. V.

2017-12-01

It is shown that a carbide-free bainite structure can be formed in high-carbon steel of the Fe-Si-Mn-Cr-V system using a traditional furnace facility. The structural aspects of bainitic transformation developing under isothermal conditions at 300°C have been studied by the methods of X-ray diffraction and transmission electron microscopy. Orientation relationships between crystalline lattices of γ and α phases have been established. A superequilibrium carbon concentration in the bainite α phase has been determined.

Making sense of nanocrystal lattice fringes

International Nuclear Information System (INIS)

Fraundorf, P.; Qin Wentao; Moeck, Peter; Mandell, Eric

2005-01-01

The orientation dependence of thin-crystal lattice fringes can be gracefully quantified using fringe-visibility maps, a direct-space analog of Kikuchi maps [Nishikawa and Kikuchi, Nature (London) 121, 1019 (1928)]. As in navigation of reciprocal space with the aid of Kikuchi lines, fringe-visibility maps facilitate acquisition of crystallographic information from lattice images. In particular, these maps can help researchers to determine the three-dimensional lattice of individual nanocrystals, to 'fringe-fingerprint' collections of randomly oriented particles, and to measure local specimen thickness with only a modest tilt. Since the number of fringes in an image increases with maximum spatial-frequency squared, these strategies (with help from more precise goniometers) will be more useful as aberration correction moves resolutions into the subangstrom range

Fuel-coolant interaction-phenomena under prompt burst conditions

International Nuclear Information System (INIS)

Jacobs, H.; Young, M.F.; Reil, K.O.

1979-01-01

The Prompt Burst Energetics (PBE) experiments conducted at Sandia Laboratories are a series of in-pile tests with fresh uranium oxide or uranium carbide fuel pins in stagnant sodium. Fuel-coolant-interactions in PBE-9S (oxide/sodium system) and PBE-SG2 (carbide/sodium) have been analyzed with the MURTI parametric FCI code. The purpose is to gain insight into possible FCI scenarios in the experiments and sensitivity of results to input parameters. Results are in approximate agreement for the second (triggered) event in PBE-9S (32 MPa peak) and the initial interaction in PBE-SG2

Renormalization group treatment of bond percolation in anisotropic and 'inhomogeneous' planar lattices

International Nuclear Information System (INIS)

Magalhaes, A.C.N. de; Tsallis, C.; Schwaccheim, G.

1980-04-01

The uncorrelated bond percolation problem is studied in three planar systems where there are two distinct occupancy probabilities. Two different real space renormalization group approaches (referred as the 'canonical' (CRG) and the 'parametric' (PRG) ones) are applied to the anisotropic first-neighbour square lattice, and both of them exhibit the expected tendency towards the exactly known phase boundary (p+q=1). Then, within the context of PRG calculations for increasingly large cells, an extrapolation method is introduced, which leads to analytic proposals for the other two lattices, namely p+q = 1/2 for the first-and second-neighbour square lattice (p and q are, respectively, the first and second neighbour occupancy probabilities), and 3 (p-1/2) = 4 [(1-q) 2 + (1-q) 3 ] (p and q are, respectively, the occupancy probabilities of the topologically different bonds which are in a 1:2 ratio) for the 4- 8 lattice. (Author) [pt

Microstructural Study of Titanium Carbide Coating on Cemented Carbide

DEFF Research Database (Denmark)

Vuorinen, S.; Horsewell, Andy

1982-01-01

Titanium carbide coating layers on cemented carbide substrates have been investigated by transmission electron microscopy. Microstructural variations within the typically 5µm thick chemical vapour deposited TiC coatings were found to vary with deposit thickness such that a layer structure could...... be delineated. Close to the interface further microstructural inhomogeneities were obsered, there being a clear dependence of TiC deposition mechanism on the chemical and crystallographic nature of the upper layers of the multiphase substrate....

Equation of states for the infinite cluster and backbone in anisotropic square lattice

International Nuclear Information System (INIS)

Silva, L.R. da; Almeida, N.S.; Tsallis, C.

1985-01-01

A real space renormalization group procedure recently developed for calculating equations of states for geometrical problems is used, to treat bond percolation in the anisotropic square lattice. By choosing a convenient self-dual cluster, for all values of the occupancy probabilities P sub(x) and P sub(y) (along the x and y axes respectively), the order parameters P infinity (P sub(x),P sub(y)) and P sup(B) infinity (P sub(x),P sub(y)) respectively associated with the complete percolating infinite cluster and with its backbone are calculated. An interesting difference appears between these two quantities whenever one of the occupancy probabilities, say P sub(y), equals unity: lim sub(P sub(y) → l) P infinity (P sub(x),P sub(y) is discontinuous at P sub(x)=0 (where P sub(infinity) jumps from 0 to 1), whereas lim sub(P sub(y) → 1) P sup(B) sub(infinity) (P sub(x),P sub(y)) continuously increases from 0 to 1 when P sub(x) increases from 0 to 1. Through a convenient extrapolation procedure which includes the use of the best available values for the critical exponents β and β sup(B), values for P sub(infinity) and P sup(B) sub(infinity) which are believed to be numerically quite reliable are obtained. In particular, P sub(infinity) (p,p) approx. A (p-1/2) sup(β) (β=5/36 and A approx. 1.25) and P sup(B) sub(infinity) (p,p) approx. A sup(B) (p-1/2) sup(β) sup(B) (β sup(B) approx. 0.53 and A sup(B) approx. 1.92). (Author) [pt

Size-scaling of tensile failure stress in boron carbide

Energy Technology Data Exchange (ETDEWEB)

Wereszczak, Andrew A [ORNL; Kirkland, Timothy Philip [ORNL; Strong, Kevin T [ORNL; Jadaan, Osama M. [University of Wisconsin, Platteville; Thompson, G. A. [U.S. Army Dental and Trauma Research Detachment, Greak Lakes

2010-01-01

Weibull strength-size-scaling in a rotary-ground, hot-pressed boron carbide is described when strength test coupons sampled effective areas from the very small (~ 0.001 square millimeters) to the very large (~ 40,000 square millimeters). Equibiaxial flexure and Hertzian testing were used for the strength testing. Characteristic strengths for several different specimen geometries are analyzed as a function of effective area. Characteristic strength was found to substantially increase with decreased effective area, and exhibited a bilinear relationship. Machining damage limited strength as measured with equibiaxial flexure testing for effective areas greater than ~ 1 mm2 and microstructural-scale flaws limited strength for effective areas less than 0.1 mm2 for the Hertzian testing. The selections of a ceramic strength to account for ballistically-induced tile deflection and to account for expanding cavity modeling are considered in context with the measured strength-size-scaling.

Criticality analysis of the CAREM-25 reactor irradiated fuel elements storage pool

International Nuclear Information System (INIS)

Albornoz, A.F.; Jatuff, F.E.; Gho, C.J.

1993-01-01

A criticality safety analysis of the irradiated fuel element pool storage of the CAREM-25 reactor was performed. The CAREM project is property of the Comision Nacional de Energia Atomica (CNEA) of Argentine, and it is being executed by INVAP S.E. difficult evaluation of the CAREM core (relatively high -3,4%- enriched U O 2 , Gd 2 O 3 burnable absorber in different densities, or criticality achievement with as few as 7 fuel elements is inherited by the pool storage. The lattice code CONDOR 1.1 was used for investigating the problem scene, and some results compared on the Monte Carlo codes MONK 5.0 and MONK 6.3. Circular and square tubes of 304-L stainless steel, borated steel and boral B 4 C in Al) were tested as suitable channels for fuel element containment, in square and hexagonal arrays; in addition, burnup, burnable absorber concentration, Sm and leakage credits were determined. It was found that the critical is strongly dependent on the separation of the fuel elements in the pool. Out-of-nominal conditions were investigated too, showing that the loss of coolant and the change in temperature and density conditions in the storage lead to an increase in reactivity, but the system's reactivity remains near the safety limits. (author)

Study of diffusion of wave packets in a square lattice under external fields along the discrete nonlinear Schrödinger equation

International Nuclear Information System (INIS)

Brito, P.E. de; Nazareno, H.N.

2012-01-01

The object of the present work is to analyze the effect of nonlinearity on wave packet propagation in a square lattice subject to a magnetic and an electric field in the Hall configuration, by using the Discrete Nonlinear Schrödinger Equation (DNLSE). In previous works we have shown that without the nonlinear term, the presence of the magnetic field induces the formation of vortices that remain stationary, while a wave packet is introduced in the system. As for the effect of an applied electric field, it was shown that the vortices propagate in a direction perpendicular to the electric field, similar behavior as presented in the classical treatment, we provide a quantum mechanics explanation for that. We have performed the calculations considering first the action of the magnetic field as well as the nonlinearity. The results indicate that for low values of the nonlinear parameter U the vortices remain stationary while preserving the form. For greater values of the parameter the picture gets distorted, the more so, the greater the nonlinearity. As for the inclusion of the electric field, we note that for small U, the wave packet propagates perpendicular to the applied field, until for greater values of U the wave gets partially localized in a definite region of the lattice. That is, for strong nonlinearity the wave packet gets partially trapped, while the tail of it can propagate through the lattice. Note that this tail propagation is responsible for the over-diffusion for long times of the wave packet under the action of an electric field. We have produced short films that show clearly the time evolution of the wave packet, which can add to the understanding of the dynamics.

Neutron measurements of the vortex lattice in YBa2Cu3O7

International Nuclear Information System (INIS)

Mook, H.A.; Yethiraj, M.; Wignall, G.D.; Forgan, E.M.; Lee, S.L.; Cubitt, R.; McK. Paul, D.; Armstrong, T.

1992-01-01

Neutron diffraction has been used to measure the vortex lattice scattering for YBa 2 Cu 3 O 7 . A square pattern is found when the field is applied along the c axis, while a triangular pattern is found when the field is applied well away from the c axis. High-resolution measurements for the square pattern show that the vortex lattice has long-range orientational order but only short-range positional order. The temperature dependence of the penetration depth is not that expected for a superconductor with a conventional s-wave BCS type gap. Preliminary measurements for temperatures near the irreversibility line are consistent with the occurrence of melting of the vortex lattice or glass phase

Dismantling method for nuclear fuel assembly

International Nuclear Information System (INIS)

Yamazaki, Shuji; Kato, Akihiro; Yoshida, Masafumi.

1993-01-01

An upper nozzle is detached from a control rod guide tube and an instrumentation tube. Subsequently, slots (slits) having a predetermined width are formed longitudinally at enlarged diameter portions of the control rod guide tube and the instrumentation tube. Then, the control rod guide tube and the instrumentation tube are separated from a lower nozzle, and pulled out from the lattice space of each of the support lattices. Thereafter, a predetermined key is inserted to a key insertion window formed at each of the support lattices, to distort a spring and take the fuel rod out of the lattice space of each of the support lattices. With such procedures, when the control rod guide tube and the instrumentation tube are pulled out of the lattice space of the support lattice, the enlarged diameter portion is narrowed to reduce the diameter, thereby enabling to take them out easily. Accordingly, since the space for inserting the key can be ensured, the nuclear fuel assemblies can easily be dismantled. In addition, fuel rods can be taken out smoothly and in an intact state. (I.N.)

Hybrid SN Laplace Transform Method For Slab Lattice Calculations

International Nuclear Information System (INIS)

Segatto, Cynthia F.; Vilhena, Marco T.; Zani, Jose H.; Barros, Ricardo C.

2008-01-01

In typical lattice cells where a highly absorbing, small fuel element is embedded in the moderator, a large weakly absorbing medium, high-order transport methods become unnecessary. In this paper we describe a hybrid discrete ordinates (S N ) method for slab lattice calculations. This hybrid S N method combines the convenience of a low-order S N method in the moderator with a high-order S N method in the fuel. We use special fuel-moderator interface conditions based on an approximate angular flux interpolation analytical method and the Laplace transform (LTS N ) numerical method to calculate the neutron flux distribution and the thermal disadvantage factor. We present numerical results for a range of typical model problems. (authors)

Influence of defects on the effective electrical conductivity of a monolayer produced by random sequential adsorption of linear k-mers onto a square lattice

Science.gov (United States)

Tarasevich, Yuri Yu.; Laptev, Valeri V.; Goltseva, Valeria A.; Lebovka, Nikolai I.

2017-07-01

The effect of defects on the behaviour of electrical conductivity, σ, in a monolayer produced by the random sequential adsorption of linear k-mers (particles occupying k adjacent sites) onto a square lattice is studied by means of a Monte Carlo simulation. The k-mers are deposited on the substrate until a jamming state is reached. The presence of defects in the lattice (impurities) and of defects in the k-mers with concentrations of dl and dk, respectively, is assumed. The defects in the lattice are distributed randomly before deposition and these lattice sites are forbidden for the deposition of k-mers. The defects of the k-mers are distributed randomly on the deposited k-mers. The sites filled with k-mers have high electrical conductivity, σk, whereas the empty sites, and the sites filled by either types of defect have a low electrical conductivity, σl, i.e., a high-contrast, σk /σl ≫ 1, is assumed. We examined isotropic (both the possible x and y orientations of a particle are equiprobable) and anisotropic (all particles are aligned along one given direction, y) deposition. To calculate the effective electrical conductivity, the monolayer was presented as a random resistor network and the Frank-Lobb algorithm was used. The effects of the concentrations of defects dl and dk on the electrical conductivity for the values of k =2n, where n = 1 , 2 , … , 5, were studied. Increase of both the dl and dk parameters values resulted in decreases in the value of σ and the suppression of percolation. Moreover, for anisotropic deposition the electrical conductivity along the y direction was noticeably larger than in the perpendicular direction, x. Phase diagrams in the (dl ,dk)-plane for different values of k were obtained.

Improved spin wave theory: An application to the spin-1/2 antiferromagnetic Heisenberg model on a square lattice

International Nuclear Information System (INIS)

Tao, Ruibao.

1991-09-01

A method is developed to make a Bose transformation which is restricted in proper space. A self-consistent independent spin wave representation (SCISWR) is found for two dimensional isotropic antiferromagnet of Heisenberg square lattices. In the SCISWR, we have successfully done the renormalization from both the dynamic and kinematic interaction and calculated the corrections from the correlations of the nearest neighbour and next nearest neighbour sites. An anisotropic excitation energy of spin wave in improper space is found self-consistently and has a gap. The difficulty of divergence appearing from higher order perturbation terms in the conventional spin wave theory has been overcome and the convergence in our approach seems quite good. We find the energy of ground state E approx. -0.659 in low order approximation and the magnetization of sublattice M z = 0.430 x (N/2) for system with spin 1/2. It is also proved that a physical spin excitation restricted in proper space is still isotropic and has no gap. (author). 17 refs

Contribution to the study of U-Ti and U-Pu-Ti carbides

International Nuclear Information System (INIS)

Milet, C.A.

1968-01-01

After having discussed the reasons to use (U,Pu) carbides as fast reactor fuel, we examine the influence of the addition of titanium to these carbides. A preliminary study has been done on the system of U-C-Ti and some properties have been measured such as: density, thermal expansion, electrical resistivity, atmospheric corrosion and compatibility with stainless steel. The systems U-Pu-C-Ti (Pu/U + Pu equal to 15 per cent) and U-C-Ti have been found to be very similar. There exists a two phases region (U,Pu)C + TiC, an eutectic between (U,Pu)C and TiC for approximately 15 at %. The solubilities of U + Pu in TiC and of Ti in (U,Pu)C is less than 1 % at. The addition of titanium does not markedly change thermal expansion coefficients of (U,Pu)C. However the resistance to atmospheric corrosion and compatibility with stainless steel is improved. Thermal conductivity, calculated from electrical resistivity, has increased. On the other side, the density of fissile material is lowered. The combination of (U,Pu)C + TiC seems to be the most promising alloy for application as nuclear fuel. (author) [fr

A new metal electrocatalysts supported matrix: Palladium nanoparticles supported silicon carbide nanoparticles and its application for alcohol electrooxidation

International Nuclear Information System (INIS)

Dai Hong; Chen Yanling; Lin Yanyu; Xu Guifang; Yang Caiping; Tong Yuejin; Guo Longhua; Chen Guonan

2012-01-01

In this paper, we propose a facile approach for palladium nanoparticles load using silicon carbide nanoparticles as the new supported matrix and a familiar NaBH 4 as reducer. Detailed X-ray photoelectron spectrum (XPS) and transmission electron microscopy (TEM) analysis of the resultant products indicated that palladium nanoparticles are successfully immobilized onto the surface of the silicon carbide nanoparticles with uniform size distribution between 5 and 7 nm. The relative electrochemical characterization clearly demonstrated excellent electrocatalytic activity of this material toward alcohol in alkaline electrolytes. Investigation on the characteristics of the electrocatalytic activity of this material further indicated that the palladium nanoparticles supporting on SiC are very promising for direct alcohol fuel cells (DMFCs), biosensor and electronic devices. Moreover, it was proved that silicon carbide nanoparticles with outstanding properties as support for catalysis are of strong practical interest. And the silicon carbide could perform attractive role in adsorbents, electrodes, biomedical applications, etc.

Dimers and the Critical Ising Model on lattices of genus >1

International Nuclear Information System (INIS)

Costa-Santos, Ruben; McCoy, B.M.

2002-01-01

We study the partition function of both Close-Packed Dimers and the Critical Ising Model on a square lattice embedded on a genus two surface. Using numerical and analytical methods we show that the determinants of the Kasteleyn adjacency matrices have a dependence on the boundary conditions that, for large lattice size, can be expressed in terms of genus two theta functions. The period matrix characterizing the continuum limit of the lattice is computed using a discrete holomorphic structure. These results relate in a direct way the lattice combinatorics with conformal field theory, providing new insight to the lattice regularization of conformal field theories on higher genus Riemann surfaces

Bulk diffusion in a kinetically constrained lattice gas

Science.gov (United States)

Arita, Chikashi; Krapivsky, P. L.; Mallick, Kirone

2018-03-01

In the hydrodynamic regime, the evolution of a stochastic lattice gas with symmetric hopping rules is described by a diffusion equation with density-dependent diffusion coefficient encapsulating all microscopic details of the dynamics. This diffusion coefficient is, in principle, determined by a Green-Kubo formula. In practice, even when the equilibrium properties of a lattice gas are analytically known, the diffusion coefficient cannot be computed except when a lattice gas additionally satisfies the gradient condition. We develop a procedure to systematically obtain analytical approximations for the diffusion coefficient for non-gradient lattice gases with known equilibrium. The method relies on a variational formula found by Varadhan and Spohn which is a version of the Green-Kubo formula particularly suitable for diffusive lattice gases. Restricting the variational formula to finite-dimensional sub-spaces allows one to perform the minimization and gives upper bounds for the diffusion coefficient. We apply this approach to a kinetically constrained non-gradient lattice gas in two dimensions, viz. to the Kob-Andersen model on the square lattice.

Physics study of Canada deuterium uranium lattice with coolant void reactivity analysis

Energy Technology Data Exchange (ETDEWEB)

Park, Jin Su; Lee, Hyun Suk; Tak, Tae Woo; Lee, Deok Jung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Shin, Ho Cheol [Korea Hydro and Nuclear Power Central Research Institute (KHNP-CRI), Daejeon (Korea, Republic of)

2017-02-15

This study presents a coolant void reactivity analysis of Canada Deuterium Uranium (CANDU)-6 and Advanced Canada Deuterium Uranium Reactor-700 (ACR-700) fuel lattices using a Monte Carlo code. The reactivity changes when the coolant was voided were assessed in terms of the contributions of four factors and spectrum shifts. In the case of single bundle coolant voiding, the contribution of each of the four factors in the ACR-700 lattice is large in magnitude with opposite signs, and their summation becomes a negative reactivity effect in contrast to that of the CANDU-6 lattice. Unlike the coolant voiding in a single fuel bundle, the 2 x 2 checkerboard coolant voiding in the ACR-700 lattice shows a positive reactivity effect. The neutron current between the no-void and voided bundles, and the four factors of each bundle were analyzed to figure out the mechanism of the positive coolant void reactivity of the checkerboard voiding case. Through a sensitivity study of fuel enrichment, type of burnable absorber, and moderator to fuel volume ratio, a design strategy for the CANDU reactor was suggested in order to achieve a negative coolant void reactivity even for the checkerboard voiding case.

Low temperature chemical processing of graphite-clad nuclear fuels

Science.gov (United States)

Pierce, Robert A.

2017-10-17

A reduced-temperature method for treatment of a fuel element is described. The method includes molten salt treatment of a fuel element with a nitrate salt. The nitrate salt can oxidize the outer graphite matrix of a fuel element. The method can also include reduced temperature degradation of the carbide layer of a fuel element and low temperature solubilization of the fuel in a kernel of a fuel element.

Criticality safety studies for plutonium–uranium metal fuel pin fabrication facility

International Nuclear Information System (INIS)

Stephen, Neethu Hanna; Reddy, C.P.

2013-01-01

Highlights: ► Criticality safety limits for PUMP-F facility is identified. ► The fissile mass which can be handled safely during alloy preparation is 10.5 kg. ► The number of fuel slugs which can be handled safely during injection casting is 53. ► The number of fuel slugs which can be handled safely after fuel fabrication is 71. - Abstract: This study focuses on the criticality safety during the fabrication of fast reactor metal fuel pins comprising of the fuel type U–15Pu, U–19Pu and U–19Pu–6Zr in the Plutonium–Uranium Metal fuel Pin fabrication Facility (PUMP-F). Maximum amount of fissile mass which can be handled safely during master alloy preparation, Injection casting and fuel slug preparation following fuel pin fabrication were identified and fixed based on this study. In the induction melting furnace, the fissile mass can be limited to 10.5 kg. During fuel slug preparation and fuel pin fabrication, fuel slugs and pins were arranged in hexagonal and square lattices to identify the most reactive configuration. The number of fuel slugs which can be handled safely after injection casting can be fixed to be 53, whereas after fuel fabrication it is 71

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

Science.gov (United States)

Terrani, Kurt A.

2018-04-01

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

Fuel-coolant interaction-phenomena under prompt burst conditions. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Jacobs, H.; Young, M.F.; Reil, K.O.

1979-01-01

The Prompt Burst Energetics (PBE) experiments conducted at Sandia Laboratories are a series of in-pile tests with fresh uranium oxide or uranium carbide fuel pins in stagnant sodium. Fuel-coolant-interactions in PBE-9S (oxide/sodium system) and PBE-SG2 (carbide/sodium) have been analyzed with the MURTI parametric FCI code. The purpose is to gain insight into possible FCI scenarios in the experiments and sensitivity of results to input parameters. Results are in approximate agreement for the second (triggered) event in PBE-9S (32 MPa peak) and the initial interaction in PBE-SG2 (190 MPa peak).

Analysis of possibilities for functional capacity for work rise of reactor fuel elements at nuclear engine regime

International Nuclear Information System (INIS)

Deryavko, I.I.; Perepelkin, I.G.; Pivovarov, O.S.; Storozhenko, A.N.; Tarasov, V.I.

2000-01-01

The principle results of carbide fuel rods testing during series of IVG.1 reactor starts up at regime simulating nuclear engine regime of nuclear moving power unit are given. Considerable degradation of initial fuel elements status increasing from start up to start up and which could resulted fail of separate technological channels is shown. Origin case of extreme degradation of fuel elements status are insufficient thermal strength of fuel elements operation in the field brittle state of sintered carbide material, Possible ways of artificial reinforce of fuel elements of low temperature sections, increasing its thermal strength up to required level

Madhava, Gregory, Leibnitz, and Sums of Two Squares

Indian Academy of Sciences (India)

IAS Admin

Keywords. Gregory–Leibnitz series, lattice points, sums of two squares,. Gauss circle problem. Shailesh Shirali heads the. Community Math Centre in Rishi Valley School and works in the field of teacher education. He is the author of many books and articles in mathemat- ics, written for interested students in the age range.

Nuclear reactor fuel assembly spacer grid

International Nuclear Information System (INIS)

Jabsen, F.S.

1977-01-01

A spacer grid for a nuclear fuel assembly is comprised of a lattice of grid plates forming multiple cells that are penetrated by fuel elements. Resilient protrusions and rigid protrusions projecting into the cells from the plates bear against the fuel element to effect proper support and spacing. Pairs of intersecting grid plates, disposed in a longitudinally spaced relationship, cooperate with other plates to form a lattice wherein each cell contains adjacent panels having resilient protrusions arranged opposite adjacent panels having rigid protrusions. The peripheral band bounding the lattice is provided solely with rigid protrusions projecting into the peripheral cells. (Auth.)

Fast breeder fuel cycle

International Nuclear Information System (INIS)

1978-09-01

Basic elements of the ex-reactor part of the fuel cycle (reprocessing, fabrication, waste handling and transportation) are described. Possible technical and proliferation measures are evaluated, including current methods of accountability, surveillance and protection. The reference oxide based cycle and advanced cycles based on carbide and metallic fuels are considered utilizing conventional processes; advanced nonaqueous reprocessing is also considered. This contribution provides a comprehensive data base for evaluation of proliferation risks

The Difference between Flux Spectrums of WH-type Assembly and CANDU-type Lattice

International Nuclear Information System (INIS)

Ryu, Eun Hyun; Song, Yong Mann

2014-01-01

The nuclear reactors are categorized by the material of the moderator because of its importance. The representative materials of the moderator are light water (H 2 O) and heavy water (D 2 O). Also, it is well known that the slowing-down ratio of D 2 O is hundreds of times larger than that of H 2 O while the slowing-down power of H 2 O is several times larger than that of D 2 O. This means that the H 2 O sometimes plays a role of an absorber such as the liquid zone controller (LZC) in a CANDU-type reactor. It is thought that the flux spectrums in a different reactor can differ from each other. In this research, two representative assemblies (the Westinghouse (WH)-type fuel assembly of PWR and the CANDU-type fuel lattice of PHWR) are selected and the flux results for each group are extracted. Although there are many codes for the lattice transport calculation, the WIMS code and the HELIOS code are used for the calculation of the WH-type fuel lattice and the CANDU-type fuel lattice. A clear difference in spectrum between the CANDU-type lattice and WH 16GD-type lattice is confirmed. Because of the superior moderating ratio of the heavy water, the thermal flux ratio of the CANDU-type lattice is almost 82%, while that of the WH 16 GD-type lattice is around 23%. Because of the large portion of the thermal flux in the CANDU-type lattice, the boron effect is maximized with the result from variations of boron. Thus it can be said that the spectrum largely depends on the moderator material, and the boron effect and sensitivity largely depends on the flux spectrum. Because of the dominant effect of the moderator material on the flux spectrum in a nuclear reactor, in the future, a comparison of the spectra of SFR, HTGR, PWR, and PHWR are also an interesting subject to study. Over-moderation in PHWR lattice and under-moderation in PWR lattice can be explained by the investigation about flux spectrums with variations of moderator density in each lattice

The Difference between Flux Spectrums of WH-type Assembly and CANDU-type Lattice

Energy Technology Data Exchange (ETDEWEB)

Ryu, Eun Hyun; Song, Yong Mann [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

The nuclear reactors are categorized by the material of the moderator because of its importance. The representative materials of the moderator are light water (H{sub 2}O) and heavy water (D{sub 2}O). Also, it is well known that the slowing-down ratio of D{sub 2}O is hundreds of times larger than that of H{sub 2}O while the slowing-down power of H{sub 2}O is several times larger than that of D{sub 2}O. This means that the H{sub 2}O sometimes plays a role of an absorber such as the liquid zone controller (LZC) in a CANDU-type reactor. It is thought that the flux spectrums in a different reactor can differ from each other. In this research, two representative assemblies (the Westinghouse (WH)-type fuel assembly of PWR and the CANDU-type fuel lattice of PHWR) are selected and the flux results for each group are extracted. Although there are many codes for the lattice transport calculation, the WIMS code and the HELIOS code are used for the calculation of the WH-type fuel lattice and the CANDU-type fuel lattice. A clear difference in spectrum between the CANDU-type lattice and WH 16GD-type lattice is confirmed. Because of the superior moderating ratio of the heavy water, the thermal flux ratio of the CANDU-type lattice is almost 82%, while that of the WH 16 GD-type lattice is around 23%. Because of the large portion of the thermal flux in the CANDU-type lattice, the boron effect is maximized with the result from variations of boron. Thus it can be said that the spectrum largely depends on the moderator material, and the boron effect and sensitivity largely depends on the flux spectrum. Because of the dominant effect of the moderator material on the flux spectrum in a nuclear reactor, in the future, a comparison of the spectra of SFR, HTGR, PWR, and PHWR are also an interesting subject to study. Over-moderation in PHWR lattice and under-moderation in PWR lattice can be explained by the investigation about flux spectrums with variations of moderator density in each

Helium diffusion in irradiated boron carbide

International Nuclear Information System (INIS)

Hollenberg, G.W.

1981-03-01

Boron carbide has been internationally adopted as the neutron absorber material in the control and safety rods of large fast breeder reactors. Its relatively large neutron capture cross section at high neutron energies provides sufficient reactivity worth with a minimum of core space. In addition, the commercial availability of boron carbide makes it attractive from a fabrication standpoint. Instrumented irradiation experiments in EBR-II have provided continuous helium release data on boron carbide at a variety of operating temperatures. Although some microstructural and compositional variations were examined in these experiments most of the boron carbide was prototypic of that used in the Fast Flux Test Facility. The density of the boron carbide pellets was approximately 92% of theoretical. The boron carbide pellets were approximately 1.0 cm in diameter and possessed average grain sizes that varied from 8 to 30 μm. Pellet centerline temperatures were continually measured during the irradiation experiments