Investigation of neutron physical features of WWER-440 assembly containing differently enriched pins and Gd burnable poison

International Nuclear Information System (INIS)

Nemes, Imre

2000-01-01

In this paper different pin-distributions of WWER-440 fuel assembly are examined. Assemblies contain 3 Gd-doped pins (Hungarian design), 6 Gd-doped pins near the assembly corners (Russian design) and differently profiled U5-enrichment in different pins. The main neutron physical characteristics of this assemblies - as the function of burnup - are calculated using HELIOS code. The calculated parameters of different assembly designs are analyzed from the standpoint of fuel cycle economy and refueling design practice. (Authors)

Benchmark criticality experiments for fast fission configuration with high enriched nuclear fuel

International Nuclear Information System (INIS)

Sikorin, S.N.; Mandzik, S.G.; Polazau, S.A.; Hryharovich, T.K.; Damarad, Y.V.; Palahina, Y.A.

2014-01-01

Benchmark criticality experiments of fast heterogeneous configuration with high enriched uranium (HEU) nuclear fuel were performed using the 'Giacint' critical assembly of the Joint Institute for Power and Nuclear Research - Sosny (JIPNR-Sosny) of the National Academy of Sciences of Belarus. The critical assembly core comprised fuel assemblies without a casing for the 34.8 mm wrench. Fuel assemblies contain 19 fuel rods of two types. The first type is metal uranium fuel rods with 90% enrichment by U-235; the second one is dioxide uranium fuel rods with 36% enrichment by U-235. The total fuel rods length is 620 mm, and the active fuel length is 500 mm. The outer fuel rods diameter is 7 mm, the wall is 0.2 mm thick, and the fuel material diameter is 6.4 mm. The clad material is stainless steel. The side radial reflector: the inner layer of beryllium, and the outer layer of stainless steel. The top and bottom axial reflectors are of stainless steel. The analysis of the experimental results obtained from these benchmark experiments by developing detailed calculation models and performing simulations for the different experiments is presented. The sensitivity of the obtained results for the material specifications and the modeling details were examined. The analyses used the MCNP and MCU computer programs. This paper presents the experimental and analytical results. (authors)

De novo assembly of a cotyledon-enriched transcriptome map of Vicia faba (L. for transfer cell research

Directory of Open Access Journals (Sweden)

Kiruba Shankari eArun Chinnappa

2015-04-01

Full Text Available Vicia faba (L. is an important cool-season grain legume species used widely in agriculture but also in plant physiology research, particularly as an experimental model to study transfer cell (TC development. Adaxial epidermal cells of isolated cotyledons can be induced to form functional TCs, thus providing a valuable experimental system to investigate genetic regulation of TC development. The genome of V. faba is exceedingly large (ca. 13 Gb, however, and limited genomic information is available for this species. To provide a resource for transcript profiling of epidermal TC development, we have undertaken de novo assembly of a cotyledon-enriched transcriptome map for V. faba. Illumina paired-end sequencing of total RNA pooled from different tissues and different stages, including isolated cotyledons induced to form TCs, generated 69.5M reads, of which 65.8M were used for assembly following trimming and quality control. Assembly using a De-Bruijn graph-based approach within CLC Genomics Workbench v6.1 generated 21,297 contigs, of which 80.6% were successfully annotated against GO terms. The assembly was validated against known V. faba cDNAs held in GenBank, including transcripts previously identified as being specifically expressed in epidermal cells across TC trans-differentiation. This cotyledon-enriched transcriptome map therefore provides a valuable tool for future transcript profiling of epidermal TC development, and also enriches the genetic resources available for this important legume crop species.

PWR fuel of high enrichment with erbia and enriched gadolinia

International Nuclear Information System (INIS)

Bejmer, Klaes-Håkan; Malm, Christian

2011-01-01

Today standard PWR fuel is licensed for operation up to 65-70 MWd/kgU, which in most cases corresponds to an enrichment of more than 5 w/o "2"3"5U. Due to criticality safety reason of storage and transportation, only fuel up to 5 w/o "2"3"5U enrichment is so far used. New fuel storage installations and transportation casks are necessary investments before the reactivity level of the fresh fuel can be significantly increased. These investments and corresponding licensing work takes time, and in the meantime a solution that requires burnable poisons in all pellets of the fresh high-enriched fuel might be used. By using very small amounts of burnable absorber in every pellet the initial reactivity can be reduced to today's levels. This study presents core calculations with fuel assemblies enriched to almost 6 w/o "2"3"5U mixed with a small amount of erbia. Some of the assemblies also contain gadolinia. The results are compared to a reference case containing assemblies with 4.95 w/o "2"3"5U without erbia, utilizing only gadolinia as burnable poison. The comparison shows that the number of fresh fuel assemblies can be reduced by 21% (which increases the batch burnup by 24%) by utilizing the erbia fuel concept. However, increased cost of uranium due to higher enrichment is not fully compensated for by the cost gain due to the reduction of the number assemblies. Hence, the fuel cycle cost becomes slightly higher for the high enrichment erbia case than for the reference case. (author)

Low-enrichment and long-life Scalable LIquid Metal cooled small Modular (SLIMM-1.2) reactor

Energy Technology Data Exchange (ETDEWEB)

El-Genk, Mohamed S., E-mail: mgenk@unm.edu [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States); Mechanical Engineering Department, University of New Mexico, Albuquerque, NM (United States); Chemical and Biological Engineering Department, University of New Mexico, Albuquerque, NM (United States); Palomino, Luis M.; Schriener, Timothy M. [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States)

2017-05-15

Beginning-Of-Life (BOL) spatial distributions of the fission power in the core at 100 MW{sub th}, and the negative temperature reactivity feedback effects. Besides decreasing the UN fuel enrichment, other parameters examined are the materials of the followers to the B{sub 4}C rods for RSS and RC, the rods in the radial blanket assemblies, and the thickness of the scalloped BeO walls of the UN fuel and radial blanket assemblies. Despite ∼13% lower fuel enrichment (15.35%) and ∼22% lower hot-clean excess reactivity ($6.29 versus $8.06 for SLIMM-1.0 at 100 MW{sub th}), the operation life of the SLIMM-1.2 is ∼6.8% longer (6.3 versus 5.9 FPY for SLIMM-1.0), and the total negative temperature reactivity feedback is slightly smaller. At BOL, the radial blanket assemblies in ring 4, and the six UN fuel assembles in ring 1 of the SLIMM-1.2 core generate 2% and 23% of the total reactor thermal power, respectively. The twelve and eighteen UN fuel assemblies in rings 2 and 3 of the SLIMM-1.2 core generate 38% and 37% of the reactor thermal power, respectively. At EOL, the thermal power generation by the UN assemblies in rings 1 and 2 of the SLIMM-1.2 core decreases to 21.5% and 35.9%, respectively, while that by the assemblies in rings 3 and 4 increases to 37.6% and 5%, respectively.

The mitochondrial genomes of Atlas Geckos (Quedenfeldtia): mitogenome assembly from transcriptomes and anchored hybrid enrichment datasets

OpenAIRE

Lyra, Mariana L.; Joger, Ulrich; Schulte, Ulrich; Slimani, Tahar; El Mouden, El Hassan; Bouazza, Abdellah; Künzel, Sven; Lemmon, Alan R.; Moriarty Lemmon, Emily; Vences, Miguel

2017-01-01

The nearly complete mitogenomes of the two species of North African Atlas geckos, Quedenfeldtia moerens and Q. trachyblepharus were assembled from anchored hybrid enrichment data and RNAseq data. Congruent assemblies were obtained for four samples included in both datasets. We recovered the 13 protein-coding genes, 22 tRNA genes, and two rRNA genes for both species, including partial control region. The order of genes agrees with that of other geckos.

Optimization of enrichment distributions in nuclear fuel assemblies loaded with Uranium and Plutonium via a modified linear programming technique

Energy Technology Data Exchange (ETDEWEB)

Cuevas Vivas, Gabriel Francisco

1999-12-01

A methodology to optimize enrichment distributions in Light Water Reactor (LWR) fuel assemblies is developed and tested. The optimization technique employed is the linear programming revised simplex method, and the fuel assembly's performance is evaluated with a neutron transport code that is also utilized in the calculation of sensitivity coefficients. The enrichment distribution optimization procedure begins from a single-value (flat) enrichment distribution until a target, maximum local power peaking factor, is achieved. The optimum rod enrichment distribution, with 1.00 for the maximum local power peaking factor and with each rod having its own enrichment, is calculated at an intermediate stage of the analysis. Later, the best locations and values for a reduced number of rod enrichments is obtained as a function of a target maximum local power peaking factor by applying sensitivity to change techniques. Finally, a shuffling process that assigns individual rod enrichments among the enrichment groups is performed. The relative rod power distribution is then slightly modified and the rod grouping redefined until the optimum configuration is attained. To verify the accuracy of the relative rod power distribution, a full computation with the neutron transport code using the optimum enrichment distribution is carried out. The results are compared and tested for assembly designs loaded with fresh Low Enriched Uranium (LEU) and plutonium Mixed Oxide (MOX) isotopics for both reactor-grade and weapons-grade plutonium were utilized to demonstrate the wide range of applicability of the optimization technique. The feature of the assembly designs used for evaluation purposes included burnable absorbers and internal water regions, and were prepared to resemble the configurations of modern assemblies utilized in commercial Boiling Water Reactor (BWRs) and Pressurized Water Reactors (PWRs). In some cases, a net improvement in the relative rod power distribution or in the

Hydrogels Based on Ag+ -Modulated Assembly of 5'-Adenosine Monophosphate for Enriching Biomolecules.

Science.gov (United States)

Hu, Yuanyuan; Xie, Dong; Wu, Yang; Lin, Nangui; Song, Aixin; Hao, Jingcheng

2017-11-07

Supramolecular hydrogels obtained by combining 5'-adenosine monophosphate (AMP) with Ag + were fabricated in this work. Their gelation capability was enhanced by increasing the concentration of Ag + or decreasing the pH. The gels are very sensitive to light, which endows them with potential applications as visible-light photosensitive materials. Coordination between the nucleobase of AMP and Ag + , as well as π-π stacking of nucleobases, are considered to be the main driving forces for self-assembly. The hydrogels successfully achieved the encapsulation and enrichment of biomolecules. Hydrogen bonding between the amino group of guest molecules and silver nanoparticles along the nanofibers drives the enrichment and is considered to be a crucial interaction. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

WWER radial reflector modeling by diffusion codes

International Nuclear Information System (INIS)

Petkov, P. T.; Mittag, S.

2005-01-01

The two commonly used approaches to describe the WWER radial reflectors in diffusion codes, by albedo on the core-reflector boundary and by a ring of diffusive assembly size nodes, are discussed. The advantages and disadvantages of the first approach are presented first, then the Koebke's equivalence theory is outlined and its implementation for the WWER radial reflectors is discussed. Results for the WWER-1000 reactor are presented. Then the boundary conditions on the outer reflector boundary are discussed. The possibility to divide the library into fuel assembly and reflector parts and to generate each library by a separate code package is discussed. Finally, the homogenization errors for rodded assemblies are presented and discussed (Author)

Calculation and analysis for a series of enriched uranium bare sphere critical assemblies

International Nuclear Information System (INIS)

Yang Shunhai

1994-12-01

The imported reactor fuel assembly MARIA program system is adapted to CYBER 825 computer in China Institute of Atomic Energy, and extensively used for a series of enriched uranium bare sphere critical assemblies. The MARIA auxiliary program of resonance modification MA is designed for taking account of the effects of resonance fission and absorption on calculated results. By which, the multigroup constants in the library attached to MARIA program are revised based on the U.S. Evaluated Nuclear Data File ENDF/B-IV, the related nuclear data files are replaced. And then, the reactor geometry buckling and multiplication factor are given in output tapes. The accuracy of calculated results is comparable with those of Monte Carlo and Sn method, and the agreement with experiment result is in 1%. (5 refs., 4 figs., 3 tabs.)

Fuel radial design using Path Relinking; Diseno radial de combustible usando Path Relinking

Energy Technology Data Exchange (ETDEWEB)

Campos S, Y. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)

2007-07-01

The present work shows the obtained results when implementing the combinatory optimization technique well-known as Path Re linking (Re-linkage of Trajectories), to the problem of the radial design of nuclear fuel assemblies, for boiling water reactors (BWR Boiling Water Reactor by its initials in English), this type of reactors is those that are used in the Laguna Verde Nucleo electric Central, Veracruz. As in any other electric power generation plant of that make use of some fuel to produce heat and that it needs each certain time (from 12 to 14 months) to make a supply of the same one, because this it wears away or it burns, in the nucleolectric plants to this activity is denominated fuel reload. In this reload different activities intervene, among those which its highlight the radial and axial designs of fuel assemblies, the patterns of control rods and the multi cycles study, each one of these stages with their own complexity. This work was limited to study in independent form the radial design, without considering the other activities. These phases are basic for the fuel reload design and of reactor operation strategies. (Author)

Experiments and Simulations of the Use of Time-Correlated Thermal Neutron Counting to Determine the Multiplication of an Assembly of Highly Enriched Uranium

Energy Technology Data Exchange (ETDEWEB)

David L. Chichester; Mathew T. Kinlaw; Scott M. Watson; Jeffrey M. Kalter; Eric C. Miller; William A. Noonan

2014-11-01

A series of experiments and numerical simulations using thermal-neutron time-correlated measurements has been performed to determine the neutron multiplication, M, of assemblies of highly enriched uranium available at Idaho National Laboratory. The experiments used up to 14.4 kg of highly-enriched uranium, including bare assemblies and assemblies reflected with high-density polyethylene, carbon steel, and tungsten. A small 252Cf source was used to initiate fission chains within the assembly. Both the experiments and the simulations used 6-channel and 8-channel detector systems, each consisting of 3He proportional counters moderated with polyethylene; data was recorded in list mode for analysis. 'True' multiplication values for each assembly were empirically derived using basic neutron production and loss values determined through simulation. A total of one-hundred and sixteen separate measurements were performed using fifty-seven unique measurement scenarios, the multiplication varied from 1.75 to 10.90. This paper presents the results of these comparisons and discusses differences among the various cases.

Uranium Enrichment Reduction in the Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR with PBO Reflector

Directory of Open Access Journals (Sweden)

Chihyung Kim

2016-04-01

Full Text Available The Korean Prototype Gen-IV sodium-cooled fast reactor (PGSFR is supposed to be loaded with a relatively-costly low-enriched U fuel, while its envisaged transuranic fuels are not available for transmutation. In this work, the U-enrichment reduction by improving the neutron economy is pursued to save the fuel cost. To improve the neutron economy of the core, a new reflector material, PbO, has been introduced to replace the conventional HT9 reflector in the current PGSFR core. Two types of PbO reflectors are considered: one is the conventional pin-type and the other one is an inverted configuration. The inverted PbO reflector design is intended to maximize the PbO volume fraction in the reflector assembly. In addition, the core radial configuration is also modified to maximize the performance of the PbO reflector. For the baseline PGSFR core with several reflector options, the U enrichment requirement has been analyzed and the fuel depletion analysis is performed to derive the equilibrium cycle parameters. The linear reactivity model is used to determine the equilibrium cycle performances of the core. Impacts of the new PbO reflectors are characterized in terms of the cycle length, neutron leakage, radial power distribution, and operational fuel cost.

Uranium enrichment reduction in the Prototype Gen-IV sodium-cooled fast reactor (PGSFR) with PBO reflector

Energy Technology Data Exchange (ETDEWEB)

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

2016-04-15

The Korean Prototype Gen-IV sodium-cooled fast reactor (PGSFR) is supposed to be loaded with a relatively-costly low-enriched U fuel, while its envisaged transuranic fuels are not available for transmutation. In this work, the U-enrichment reduction by improving the neutron economy is pursued to save the fuel cost. To improve the neutron economy of the core, a new reflector material, PbO, has been introduced to replace the conventional HT9 reflector in the current PGSFR core. Two types of PbO reflectors are considered: one is the conventional pin-type and the other one is an inverted configuration. The inverted PbO reflector design is intended to maximize the PbO volume fraction in the reflector assembly. In addition, the core radial configuration is also modified to maximize the performance of the PbO reflector. For the baseline PGSFR core with several reflector options, the U enrichment requirement has been analyzed and the fuel depletion analysis is performed to derive the equilibrium cycle parameters. The linear reactivity model is used to determine the equilibrium cycle performances of the core. Impacts of the new PbO reflectors are characterized in terms of the cycle length, neutron leakage, radial power distribution, and operational fuel cost.

Distributed collaborative probabilistic design for turbine blade-tip radial running clearance using support vector machine of regression

Science.gov (United States)

Fei, Cheng-Wei; Bai, Guang-Chen

2014-12-01

To improve the computational precision and efficiency of probabilistic design for mechanical dynamic assembly like the blade-tip radial running clearance (BTRRC) of gas turbine, a distribution collaborative probabilistic design method-based support vector machine of regression (SR)(called as DCSRM) is proposed by integrating distribution collaborative response surface method and support vector machine regression model. The mathematical model of DCSRM is established and the probabilistic design idea of DCSRM is introduced. The dynamic assembly probabilistic design of aeroengine high-pressure turbine (HPT) BTRRC is accomplished to verify the proposed DCSRM. The analysis results reveal that the optimal static blade-tip clearance of HPT is gained for designing BTRRC, and improving the performance and reliability of aeroengine. The comparison of methods shows that the DCSRM has high computational accuracy and high computational efficiency in BTRRC probabilistic analysis. The present research offers an effective way for the reliability design of mechanical dynamic assembly and enriches mechanical reliability theory and method.

Radial wedge flange clamp

Science.gov (United States)

Smith, Karl H.

2002-01-01

A radial wedge flange clamp comprising a pair of flanges each comprising a plurality of peripheral flat wedge facets having flat wedge surfaces and opposed and mating flat surfaces attached to or otherwise engaged with two elements to be joined and including a series of generally U-shaped wedge clamps each having flat wedge interior surfaces and engaging one pair of said peripheral flat wedge facets. Each of said generally U-shaped wedge clamps has in its opposing extremities apertures for the tangential insertion of bolts to apply uniform radial force to said wedge clamps when assembled about said wedge segments.

Research and design calculation of multipurpose critical assembly using moderated light water and low enriched fuel from 1.6 to 5.0% U-235

International Nuclear Information System (INIS)

Nguyen Kien Cuong; Vo Doan Hai Dang; Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Minh Tuan; Nguyen Manh Hung; Pham Quang Huy; Tran Quoc Duong; Tran Tri Vien

2015-01-01

Basing on the idea in ??using fuel of nuclear power plants such as PWR (AP-1000) and VVER-1000 with light water as moderation, design calculation of critical assembly was performed to confirm the possibility of using these fuels. Designed critical assembly has simple structure consisting of low enriched fuel from 1.6% to 5% U-235; water has functions as cooling, biological protection and control. Critical assembly is operated at nominal power 100 W with fuel pitch about 2.0 cm. Applications of the critical assembly are quite abundant in basic research, education and training with low investment cost compare with research reactor and easy in operation. So critical assembly can be used for university or training centre for nuclear engineering training. Main objectives of the project are: design calculation in neutronics, thermal hydraulics and safety analysis for critical configuration benchmarks using low enriched fuel; design in mechanical and auxiliary systems for critical assembly; determine technical specifications and estimate construction, installation cost of critical assembly. The process of design, fabrication, installation and construction of critical assembly will be considered with different implementation phases and localization capabilities in installation of critical assembly is highly feasibility. Cost estimation of construction and installation of critical assembly was implemented and showed that investment cost for critical assembly is much lower than research reactor and most of components, systems of critical assembly can be localized with current technique quality of the country. (author)

Improvements in nuclear fuel assembly sleeves

International Nuclear Information System (INIS)

Eaton, C.W.; Seeley, T.A.; Ince, G.; Speakman, W.T.

1986-01-01

The graphite sleeve of a nuclear fuel assembly or reflector element for a stringer mounts a number of grids via mounting assemblies installed in grooves formed in the interior wall surface of the sleeve. The bore of the sleeve is of reduced cross-section between two successive grooves such that the internal diameter of the sleeve is substantially the same as the inner diameter of the radially innermost extremity of the mounting assemblies whereby the coolant pressure loss at each transition between the reduced diameter bore section and the mounting assemblies is reduced. Each mounting assembly may be of radially contractable split ring construction to permit its placement in the groove and may carry burnable poison material. (author)

Improvements in nuclear fuel assembly sleeves

Energy Technology Data Exchange (ETDEWEB)

Eaton, C.W.; Seeley, T.A.; Ince, G.; Speakman, W.T.

1986-02-26

The graphite sleeve of a nuclear fuel assembly or reflector element for a stringer mounts a number of grids via mounting assemblies installed in grooves formed in the interior wall surface of the sleeve. The bore of the sleeve is of reduced cross-section between two successive grooves such that the internal diameter of the sleeve is substantially the same as the inner diameter of the radially innermost extremity of the mounting assemblies whereby the coolant pressure loss at each transition between the reduced diameter bore section and the mounting assemblies is reduced. Each mounting assembly may be of radially contractable split ring construction to permit its placement in the groove and may carry burnable poison material.

Improvements in nuclear fuel assembly cages

Energy Technology Data Exchange (ETDEWEB)

Eaton, C.W.; Seeley, T.A.; Ince, G.; Speakman, W.T.

1986-03-12

The fuel pin/guide tube supporting grids of an assembly cage for a multi pin fuel element or a reflector element for a stringer are mounted in the moderator sleeve by way of mounting assemblies engaged in grooves machined into the interior surface of the sleeve, each mounting assembly including a split ring which is assembled into its groove by being radially contracted, pushed along the sleeve into registry with the groove and allowed to radially expand. The split ring may carry burnable neutron absorber. The region of the sleeve between two adjacent grids may be of smaller internal diameter than the remainder of the sleeve.

Fuel radial design using Path Relinking

International Nuclear Information System (INIS)

Campos S, Y.

2007-01-01

The present work shows the obtained results when implementing the combinatory optimization technique well-known as Path Re linking (Re-linkage of Trajectories), to the problem of the radial design of nuclear fuel assemblies, for boiling water reactors (BWR Boiling Water Reactor by its initials in English), this type of reactors is those that are used in the Laguna Verde Nucleo electric Central, Veracruz. As in any other electric power generation plant of that make use of some fuel to produce heat and that it needs each certain time (from 12 to 14 months) to make a supply of the same one, because this it wears away or it burns, in the nucleolectric plants to this activity is denominated fuel reload. In this reload different activities intervene, among those which its highlight the radial and axial designs of fuel assemblies, the patterns of control rods and the multi cycles study, each one of these stages with their own complexity. This work was limited to study in independent form the radial design, without considering the other activities. These phases are basic for the fuel reload design and of reactor operation strategies. (Author)

Intra-assembly flow redistribution in LMFBRs: a simple computational approach

International Nuclear Information System (INIS)

Khatib-Rahbar, M.; Cazzoli, E.G.

1983-01-01

The liquid metal fast breeder reactor (LMFBR) core consists of fuel, blanket, control, and shielding assemblies packed in a hexagonal configuration. Radial blanket assemblies occupy peripheral locations in the reactor core and are characterized by steep power gradients, while inner blanket assemblies are located within the fuel assembly region and have higher power levels but flatter distributions. It is due to the presence of this radial power gradient that large sodium temperature distributions exist at full power operation. However, at low power, low flow natural convection conditions, a significant flow redistribution takes place leading to considerable radial temperature flattening. The purpose of the present study is to formulate a simple flow-regime dependent model supported by experimental data for prediction of sodium temperature flattening due to buoyancy-induced flow redistribution in LMFBR subassemblies with significant radial power gradient

Simulated nuclear reactor fuel assembly

International Nuclear Information System (INIS)

Berta, V.T.

1993-01-01

An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end

Sensor mount assemblies and sensor assemblies

Science.gov (United States)

Miller, David H [Redondo Beach, CA

2012-04-10

Sensor mount assemblies and sensor assemblies are provided. In an embodiment, by way of example only, a sensor mount assembly includes a busbar, a main body, a backing surface, and a first finger. The busbar has a first end and a second end. The main body is overmolded onto the busbar. The backing surface extends radially outwardly relative to the main body. The first finger extends axially from the backing surface, and the first finger has a first end, a second end, and a tooth. The first end of the first finger is disposed on the backing surface, and the tooth is formed on the second end of the first finger.

Study of Fuel Rods Axial Enrichment Distribution Effect on the Neutronic Parameters of the Reactor Core

International Nuclear Information System (INIS)

Pazirandeh, A.; Nasiri, S. H.

2012-01-01

Optimization of the fuel burn up is an important issue in nuclear reactor fuel management and technology. Radial enrichment distribution in the reactor core is a conventional method and axial enrichment is constant along the fuel rod. In this article, the effects of axial enrichment distribution variation on neutronic parameters of PWR core are studied. The axial length of the core is divided into ten sections, considering axial enrichment variation and leaving the existing radial enrichment distribution intact. This study shows that the radial and axial power peaking factors are decreased as compared with the typical conventional core. In addition, the first core lifetime lasts 30 days longer than normal PWR core. Moreover, at the same time boric acid density is 0.2 g/kg at the beginning of the cycle. The flux shape is also flat at the beginning of the cycle for the proposed configuration of the axially enrichment distribution.

Nuclear reactor spacer assembly

International Nuclear Information System (INIS)

Anthony, A.J.; Groves, M.D.

1979-01-01

A fuel assembly for a nuclear reactor is disclosed wherein the fuel element receiving and supporting grid is comprised of a first metal, the guide tubes which pass through the grid assembly are comprised of a second metal and the grid is supported on the guide tubes by means of expanded sleeves located intermediate the grid and guide tubes. The fuel assembly is fabricated by inserting the sleeves, of initial outer diameter commensurate with the guide tube outer diameters, through the holes in the grid assembly provided for the guide tubes and thereafter expanding the sleeves radially outwardly along their entire length such that the guide tubes can subsequently be passed through the sleeves. The step of radial expansion, as a result of windows provided in the sleeves having dimensions commensurate with the geometry of the grid, mechanically captures the grid and simultaneously preloads the sleeve against the grid whereby relative motion between the grid and guide tube will be precluded

ALARA (As Low As Reasonable Achievable) procedure applied to fuel assembly fabrication with enriched reprocessing uranium (ERU)

International Nuclear Information System (INIS)

Guimaraes, Leonam dos Santos; Degrange, Jean Pierre

1998-01-01

The study introduced by this paper compose the first step to the implementation of ALARA (As Low As Reasonable Achievable) for a nuclear fuel assembly factory which one of its two production lines will be designed to work with Enriched Reprocessing Uranium (ERU). This step includes the reference situation analysis is based on previsional dosimetric evaluations for individual and collective exposures of each factory operator (117 in total) working on 7 work stations, considering 6 annual production scenarios (10, 50 75, 100 and 150 ERU tons), which corresponds to an annual production of 600 tons (ERU plus enriched natural uranium ENU). The exposure indicators evolution, expressed in terms of collective dose, annual individual dose and radiological detrimental cost for workers, is also used in a complimentary way to guide the analysis. (author)

The radial distribution of plutonium in high burnup UO2 fuels

International Nuclear Information System (INIS)

Lassmann, K.; O'Carroll, C.; Laar, J. van de; Walker, C.T.

1994-01-01

A new model (TUBRNP) is described which predicts the radial power density distribution as a function of burnup (and hence the radial burnup profile as a function of time) together with the radial profile of uranium and plutonium isotopes. Comparisons between measurements and the predictions of the TUBRNP model are made on fuels with enrichments in the range 2.9 to 8.25% and with burnups between 21 000 and 64 000 MWd/t. It is shown to be in excellent agreement with experimental measurements and is a marked improvement on earlier versions. (orig.)

The innate origin of radial and vertical gradients in a simulated galaxy disc

Science.gov (United States)

Navarro, Julio F.; Yozin, Cameron; Loewen, Nic; Benítez-Llambay, Alejandro; Fattahi, Azadeh; Frenk, Carlos S.; Oman, Kyle A.; Schaye, Joop; Theuns, Tom

2018-05-01

We examine the origin of radial and vertical gradients in the age/metallicity of the stellar component of a galaxy disc formed in the APOSTLE cosmological hydrodynamical simulations. Some of these gradients resemble those in the Milky Way, where they have sometimes been interpreted as due to internal evolution, such as scattering off giant molecular clouds, radial migration driven by spiral patterns, or orbital resonances with a bar. Secular processes play a minor role in the simulated galaxy, which lacks strong spiral or bar patterns, and where such gradients arise as a result of the gradual enrichment of a gaseous disc that is born thick but thins as it turns into stars and settles into centrifugal equilibrium. The settling is controlled by the feedback of young stars; which links the star formation, enrichment, and equilibration time-scales, inducing radial and vertical gradients in the gaseous disc and its descendent stars. The kinematics of coeval stars evolve little after birth and provide a faithful snapshot of the gaseous disc structure at the time of their formation. In this interpretation, the age-velocity dispersion relation would reflect the gradual thinning of the disc rather than the importance of secular orbit scattering; the outward flaring of stars would result from the gas disc flare rather than from radial migration; and vertical gradients would arise because the gas disc gradually thinned as it enriched. Such radial and vertical trends might just reflect the evolving properties of the parent gaseous disc, and are not necessarily the result of secular evolutionary processes.

Fuel assembly

International Nuclear Information System (INIS)

Watanabe, Shoichi; Hirano, Yasushi.

1998-01-01

A one-half or more of entire fuel rods in a fuel assembly comprises MOX fuel rods containing less than 1wt% of burnable poisons, and at least a portion of the burnable poisons comprises gadolinium. Then, surplus reactivity at an initial stage of operation cycle is controlled to eliminate burnable poisons remained unburnt at a final stage, as well as increase thermal reactivity. In addition, the content of fission plutonium is determined to greater than the content of uranium 235, and fuel rods at corner portions are made not to incorporate burnable poisons. Fuel rods not containing burnable poisons are disposed at positions in adjacent with fuel rods facing to a water rod at one or two directions. Local power at radial center of the fuel assembly is increased to flatten the distortion of radial power distribution. (N.H.)

Fast critical assembly safeguards: NDA methods for highly enriched uranium. Summary report, October 1978-September 1979

International Nuclear Information System (INIS)

Bellinger, F.O.; Winslow, G.H.

1980-12-01

Nondestructive assay (NDA) methods, principally passive gamma measurements and active neutron interrogation, have been studied for their safeguards effectiveness and programmatic impact as tools for making inventories of highly enriched uranium fast critical assembly fuel plates. It was concluded that no NDA method is the sole answer to the safeguards problem, that each of those emphasized here has its place in an integrated safeguards system, and that each has minimum facility impact. It was found that the 185-keV area, as determined with a NaI detector, was independent of highly-enriched uranium (HEU) plate irradiation history, though the random neutron driver methods used here did not permit accurate assay of irradiated plates. Containment procedures most effective for accurate assaying were considered, and a particular geometry is recommended for active interrogation by a random driver. A model, pertinent to that geometry, which relates the effects of multiplication and self-absorption, is described. Probabilities of failing to detect that plates are missing are examined

Radiation characteristics of GD-2 fuel

International Nuclear Information System (INIS)

Chrapciak, V.

2008-01-01

The assembly WWER-440 type Gd-2 has radial profile of enrichment and has 6 pins with Gd 2 O 3 . The maximal enrichment is 4.4%. Some analyses are done for assembly with flat enrichment (4.4%) and without Gd 2 O 3 . In this article are compared some characteristics (decay heat, some nuclide concentration, photon and gamma sources) for real Gd-2 assembly and for flat 4.4% assembly. The TRITON module (in SCALE 5.1) was used. (Author)

Fuel assembly

International Nuclear Information System (INIS)

Fushimi, Atsushi; Shimada, Hidemitsu; Aoyama, Motoo; Nakajima, Junjiro

1998-01-01

In a fuel assembly for an n x n lattice-like BWR type reactor, n is determined to 9 or greater, and the enrichment degree of plutonium is determined to 4.4% by weight or less. Alternatively, n is determined to 10 or greater, and the enrichment degree of plutonium is determined to 5.2% by weight or less. An average take-out burnup degree is determined to 39GWd/t or less, and the matrix is determined to 9 x 9 or more, or the average take-out burnup degree is determined to 51GWd/t, and the matrix is determined to 10 x 10 or more and the increase of the margin of the maximum power density obtained thereby is utilized for the compensation of the increase of distortion of power distribution due to decrease of the kinds of plutonium enrichment degree, thereby enabling to reduce the kind of the enrichment degree of MOX fuel rods to one. As a result, the manufacturing step for fuel pellets can be simplified to reduce the manufacturing cost for MOX fuel assemblies. (N.H.)

Experimental validation of CASMO-4E and CASMO-5M for radial fission rate distributions in a westinghouse SVEA-96 Optima2 BWR fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Grimm, P.; Perret, G. [Paul Scherrer Inst., CH-5232 Villigen PSI (Switzerland)

2012-07-01

Measured and calculated radial total fission rate distributions are compared for the three axial sections of a Westinghouse SVEA-96 Optima2 BWR fuel assembly, comprising 96, 92 and 84 fuel rods, respectively. The measurements were performed on a full-size fuel assembly in the PROTEUS zero-power experimental facility. The measured fission rates are compared to the results of the CASMO-4E and CASMO-5M fuel assembly codes. Detailed measured geometrical data were used in the models, and effects of the surrounding zones of the reactor were taken into account by correction factors derived from MCNPX calculations. The results of the calculations agree well with those of the experiments, with root-mean-square deviations between 1.2% and 1.5% and maximum deviations of 3-4%. The quality of the predictions by CASMO-4E and CASMO-5M is comparable. (authors)

Fission rate distribution at the 84-pin radial section of a SVEA-96 Optima2 BWR assembly

Energy Technology Data Exchange (ETDEWEB)

Perret, Gregory; Murphy, Michael F.; Jatuff, Fabian [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Chawla, Rakesh [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne (EPFL), CH-1015 Lausanne (Switzerland)

2008-07-01

Westinghouse boiling water reactor SVEA-96 Optima2 assemblies were studied during the LWRPROTEUS program at the PROTEUS facility in the Paul Scherrer Institute. Measured radial fission rate distributions at the 84-pin elevation are compared with MCNPX predictions using both ENDF/B-VI (Release 2) and JEFF-3.1 data libraries. Predicted fission rates agree within +-4.5% using both libraries. Fission rates were over-predicted in UO{sub 2} pins close to the missing 1/3 pins and under-predicted in UO{sub 2} pins close to the missing 2/3 pins. Recurrent under-estimations were observed in the UO{sub 2}-Gd{sub 2}O{sub 3} pins, for both libraries, which might be explained by over-estimated thermal cross-sections of {sup 157}Gd, as suggested in a recent work of G. Leinweber et al. (2006). (authors)

Contemporary and prospective fuel cycles for WWER-440 based on new assemblies with higher uranium capacity and higher average fuel enrichment

International Nuclear Information System (INIS)

Gagarinskiy, A.A.; Saprykin, V.V.

2009-01-01

RRC 'Kurchatov Institute' has performed an extensive cycle of calculations intended to validate the opportunities of improving different fuel cycles for WWER-440 reactors. Works were performed to upgrade and improve WWER-440 fuel cycles on the basis of second-generation fuel assemblies allowing core thermal power to be uprated to 107 108 % of its nominal value (1375 MW), while maintaining the same fuel operation lifetime. Currently intensive work is underway to develop fuel cycles based on second-generation assemblies with higher fuel capacity and average fuel enrichment per assembly increased up to 4.87 % of U-235. Fuel capacity of second-generation assemblies was increased by means of eliminated central apertures of fuel pellets, and pellet diameter extended due to reduced fuel cladding thickness. This paper intends to summarize the results of works performed in the field of WWER-440 fuel cycle modernization, and to present yet unemployed opportunities and prospects of further improvement of WWER-440 neutronic and operating parameters by means of additional optimization of fuel assembly designs and fuel element arrangements applied. (Authors)

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

Reactor core and control rod assembly in FBR type reactor

International Nuclear Information System (INIS)

Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi.

1993-01-01

Fuel assemblies and control rod assemblies are attached respectively to reactor core support plates each in a cantilever fashion. Intermediate spacer pads are disposed to the lateral side of a wrapper tube just above the fuel rod region. Intermediate space pads are disposed to the lateral side of a control rod guide tube just above a fuel rod region. The thickness of the intermediate spacer pad for the control rod assembly is made smaller than the thickness of the intermediate spacer pad for the fuel assembly. This can prevent contact between intermediate spacer pads of the control guide tube and the fuel assembly even if the temperature of coolants is elevated to thermally expand the intermediate spacer pad, by which the radial displacement amount of the reactor core region along the direction of the height of the control guide tube is reduced substantially to zero. Accordingly, contribution of the control rod assembly to the radial expansion reactivity can be reduced to zero or negative level, by which the effect of the negative radial expansion reactivity of the reactor is increased to improve the safety upon thermal transient stage, for example, loss of coolant flow rate accident. (I.N.)

Radial flow heat exchanger

Science.gov (United States)

Valenzuela, Javier

2001-01-01

A radial flow heat exchanger (20) having a plurality of first passages (24) for transporting a first fluid (25) and a plurality of second passages (26) for transporting a second fluid (27). The first and second passages are arranged in stacked, alternating relationship, are separated from one another by relatively thin plates (30) and (32), and surround a central axis (22). The thickness of the first and second passages are selected so that the first and second fluids, respectively, are transported with laminar flow through the passages. To enhance thermal energy transfer between first and second passages, the latter are arranged so each first passage is in thermal communication with an associated second passage along substantially its entire length, and vice versa with respect to the second passages. The heat exchangers may be stacked to achieve a modular heat exchange assembly (300). Certain heat exchangers in the assembly may be designed slightly differently than other heat exchangers to address changes in fluid properties during transport through the heat exchanger, so as to enhance overall thermal effectiveness of the assembly.

Reflector-moderated critical assemblies

International Nuclear Information System (INIS)

Paxton, H.C.; Jarvis, G.A.; Byers, C.C.

1975-07-01

Experiments with reflector-moderated critical assemblies were part of the Rover Program at the Los Alamos Scientific Laboratory (LASL). These assemblies were characterized by thick D 2 O or beryllium reflectors surrounding large cavities that contained highly enriched uranium at low average densities. Because interest in this type of system has been revived by LASL Plasma Cavity Assembly studies, more detailed descriptions of the early assemblies than had been available in the unclassified literature are provided. (U.S.)

Top-nozzle mounted replacement guide pin assemblies

International Nuclear Information System (INIS)

Gilmore, C.B.; Andrews, W.H.

1993-01-01

A replacement guide pin assembly is provided for aligning a nuclear fuel assembly with an upper core plate of a nuclear reactor core. The guide pin assembly includes a guide pin body having a radially expandable base insertable within a hole in the top nozzle, a ferrule insertable within the guide pin base and capable of imparting a radially and outwardly directed force on the expandable base to expand it within the hole of the top nozzle and thereby secure the guide pin body to the top nozzle in response to a predetermined displacement of the ferrule relative to the guide pin body along its longitudinal axis, and a lock screw interfitted with the ferrule and threaded into the guide pin body so as to produce the predetermined displacement of the ferrule. (author)

Mixer Assembly for a Gas Turbine Engine

Science.gov (United States)

Dai, Zhongtao (Inventor); Cohen, Jeffrey M. (Inventor); Fotache, Catalin G. (Inventor); Smith, Lance L. (Inventor); Hautman, Donald J. (Inventor)

2018-01-01

A mixer assembly for a gas turbine engine is provided, including a main mixer with fuel injection holes located between at least one radial swirler and at least one axial swirler, wherein the fuel injected into the main mixer is atomized and dispersed by the air flowing through the radial swirler and the axial swirler.

Shaft seal assembly for high speed and high pressure applications

Science.gov (United States)

Hadt, W. F.; Ludwig, L. P. (Inventor)

1979-01-01

A seal assembly is provided for reducing the escape of fluids from between a housing and a shaft rotably mounted in the housing. The seal assembly comprises a pair of seal rings resiliently connected to each other and disposed in side-by-side relationship. In each seal ring, both the internal bore surface and the radial face which faces away from the other seal ring are provided with a plurality of equi-spaced recesses. The seal faces referred to are located adjacent a seating surface of the housing. Under normal operating conditions, the seal assembly is stationary with respect to the housing, and the recesses generate life, keep the assembly spaced from the rotating shaft and allow slip therebetween. The seal assembly can seize on the shaft, and slip will then occur between the radial faces and the housing.

Hyb-Seq: Combining Target Enrichment and Genome Skimming for Plant Phylogenomics

Directory of Open Access Journals (Sweden)

Kevin Weitemier

2014-08-01

Full Text Available Premise of the study: Hyb-Seq, the combination of target enrichment and genome skimming, allows simultaneous data collection for low-copy nuclear genes and high-copy genomic targets for plant systematics and evolution studies. Methods and Results: Genome and transcriptome assemblies for milkweed (Asclepias syriaca were used to design enrichment probes for 3385 exons from 768 genes (>1.6 Mbp followed by Illumina sequencing of enriched libraries. Hyb-Seq of 12 individuals (10 Asclepias species and two related genera resulted in at least partial assembly of 92.6% of exons and 99.7% of genes and an average assembly length >2 Mbp. Importantly, complete plastomes and nuclear ribosomal DNA cistrons were assembled using off-target reads. Phylogenomic analyses demonstrated signal conflict between genomes. Conclusions: The Hyb-Seq approach enables targeted sequencing of thousands of low-copy nuclear exons and flanking regions, as well as genome skimming of high-copy repeats and organellar genomes, to efficiently produce genome-scale data sets for phylogenomics.

Leakage Account for Radial Face Contact Seal in Aircraft Engine Support

Science.gov (United States)

Vinogradov, A. S.; Sergeeva, T. V.

2018-01-01

The article is dedicated to the development of a methodology for the radial face contact seal design taking into consideration the supporting elements deformations in different aircraft engine operating modes. Radial face contact seals are popular in the aircraft engines bearing support. However, there are no published leakage calculation methodologies of these seals. Radial face contact seal leakage is determined by the gap clearance in the carbon seal ring split. In turn, the size gap clearance depends on the deformation of the seal assembly parts and from the engine operation. The article shows the leakage detection sequence in the intershaft radial face contact seal of the compressor support for take-off and cruising modes. Evaluated calculated leakage values (2.4 g/s at takeoff and 0.75 g/s at cruising) go with experience in designing seals.

The assembly of GM1 glycolipid- and cholesterol-enriched raft-like membrane microdomains is important for giardial encystation.

Science.gov (United States)

De Chatterjee, Atasi; Mendez, Tavis L; Roychowdhury, Sukla; Das, Siddhartha

2015-05-01

Although encystation (or cyst formation) is an important step of the life cycle of Giardia, the cellular events that trigger encystation are poorly understood. Because membrane microdomains are involved in inducing growth and differentiation in many eukaryotes, we wondered if these raft-like domains are assembled by this parasite and participate in the encystation process. Since the GM1 ganglioside is a major constituent of mammalian lipid rafts (LRs) and known to react with cholera toxin B (CTXB), we used Alexa Fluor-conjugated CTXB and GM1 antibodies to detect giardial LRs. Raft-like structures in trophozoites are located in the plasma membranes and on the periphery of ventral discs. In cysts, however, they are localized in the membranes beneath the cyst wall. Nystatin and filipin III, two cholesterol-binding agents, and oseltamivir (Tamiflu), a viral neuraminidase inhibitor, disassembled the microdomains, as evidenced by reduced staining of trophozoites with CTXB and GM1 antibodies. GM1- and cholesterol-enriched LRs were isolated from Giardia by density gradient centrifugation and found to be sensitive to nystatin and oseltamivir. The involvement of LRs in encystation could be supported by the observation that raft inhibitors interrupted the biogenesis of encystation-specific vesicles and cyst production. Furthermore, culturing of trophozoites in dialyzed medium containing fetal bovine serum (which is low in cholesterol) reduced raft assembly and encystation, which could be rescued by adding cholesterol from the outside. Our results suggest that Giardia is able to form GM1- and cholesterol-enriched lipid rafts and these raft domains are important for encystation. Copyright © 2015, American Society for Microbiology. All Rights Reserved.

Nuclear fuel string assembly

International Nuclear Information System (INIS)

Ip, A.K.; Koyanagi, K.; Tarasuk, W.R.

1976-01-01

A method of fabricating rodded fuels suitable for use in pressure tube type reactors and in pressure vessel type reactors is described. Fuel rods are secured as an inner and an outer sub-assembly, each rod attached between mounting rings secured to the rod ends. The two sub-assemblies are telescoped together and positioned by spaced thimbles located between them to provide precise positioning while permittng differential axial movement between the sub-assemblies. Such sub-assemblies are particularly suited for mounting as bundle strings. The method provides particular advantages in the assembly of annular-section fuel pins, which includes booster fuel containing enriched fuel material. (LL)

Capacitor assembly and related method of forming

Science.gov (United States)

Zhang, Lili; Tan, Daniel Qi; Sullivan, Jeffrey S.

2017-12-19

A capacitor assembly is disclosed. The capacitor assembly includes a housing. The capacitor assembly further includes a plurality of capacitors disposed within the housing. Furthermore, the capacitor assembly includes a thermally conductive article disposed about at least a portion of a capacitor body of the capacitors, and in thermal contact with the capacitor body. Moreover, the capacitor assembly also includes a heat sink disposed within the housing and in thermal contact with at least a portion of the housing and the thermally conductive article such that the heat sink is configured to remove heat from the capacitor in a radial direction of the capacitor assembly. Further, a method of forming the capacitor assembly is also presented.

Hyb-Seq: Combining target enrichment and genome skimming for plant phylogenomics1

Science.gov (United States)

Weitemier, Kevin; Straub, Shannon C. K.; Cronn, Richard C.; Fishbein, Mark; Schmickl, Roswitha; McDonnell, Angela; Liston, Aaron

2014-01-01

• Premise of the study: Hyb-Seq, the combination of target enrichment and genome skimming, allows simultaneous data collection for low-copy nuclear genes and high-copy genomic targets for plant systematics and evolution studies. • Methods and Results: Genome and transcriptome assemblies for milkweed (Asclepias syriaca) were used to design enrichment probes for 3385 exons from 768 genes (>1.6 Mbp) followed by Illumina sequencing of enriched libraries. Hyb-Seq of 12 individuals (10 Asclepias species and two related genera) resulted in at least partial assembly of 92.6% of exons and 99.7% of genes and an average assembly length >2 Mbp. Importantly, complete plastomes and nuclear ribosomal DNA cistrons were assembled using off-target reads. Phylogenomic analyses demonstrated signal conflict between genomes. • Conclusions: The Hyb-Seq approach enables targeted sequencing of thousands of low-copy nuclear exons and flanking regions, as well as genome skimming of high-copy repeats and organellar genomes, to efficiently produce genome-scale data sets for phylogenomics. PMID:25225629

Effects of Radial Reflector Composition on Core Reactivity and Peak Power

International Nuclear Information System (INIS)

Park, Sang Yoon; Lee, Kyung Hoon; Song, Jae Seung

2007-10-01

The effects of radial SA-240 alloy shroud on core reactivity and peak power are evaluated. The existence of radial SA-240 alloy shroud makes reflector water volume decrease, so the thermal absorption cross section of radial reflector is lower than without SA-240 alloy shroud case. Finally, the cycle length is increased from 788 EFPD to 845 EFPD and the peak power is decreased from 1.66 to 1.49. In the case of without SA-240 alloy shroud, a new core loading pattern search has been performed. For the guarantee of the same equivalent cycle length of with SA-240 alloy shroud case, the enrichment of U-235 should be increased from 4.22 w/o to 4.68 w/o. The nuclear key safety parameters of new core loading pattern have been calculated and recorded for the future

Effects of Radial Reflector Composition on Core Reactivity and Peak Power

Energy Technology Data Exchange (ETDEWEB)

Park, Sang Yoon; Lee, Kyung Hoon; Song, Jae Seung

2007-10-15

The effects of radial SA-240 alloy shroud on core reactivity and peak power are evaluated. The existence of radial SA-240 alloy shroud makes reflector water volume decrease, so the thermal absorption cross section of radial reflector is lower than without SA-240 alloy shroud case. Finally, the cycle length is increased from 788 EFPD to 845 EFPD and the peak power is decreased from 1.66 to 1.49. In the case of without SA-240 alloy shroud, a new core loading pattern search has been performed. For the guarantee of the same equivalent cycle length of with SA-240 alloy shroud case, the enrichment of U-235 should be increased from 4.22 w/o to 4.68 w/o. The nuclear key safety parameters of new core loading pattern have been calculated and recorded for the future.

Lateral restraint assembly for reactor core

Science.gov (United States)

Gorholt, Wilhelm; Luci, Raymond K.

1986-01-01

A restraint assembly for use in restraining lateral movement of a reactor core relative to a reactor vessel wherein a plurality of restraint assemblies are interposed between the reactor core and the reactor vessel in circumferentially spaced relation about the core. Each lateral restraint assembly includes a face plate urged against the outer periphery of the core by a plurality of compression springs which enable radial preloading of outer reflector blocks about the core and resist low-level lateral motion of the core. A fixed radial key member cooperates with each face plate in a manner enabling vertical movement of the face plate relative to the key member but restraining movement of the face plate transverse to the key member in a plane transverse to the center axis of the core. In this manner, the key members which have their axes transverse to or subtending acute angles with the direction of a high energy force tending to move the core laterally relative to the reactor vessel restrain such lateral movement.

Sodium-cooled fast reactor (SFR) fuel assembly design with graphite-moderating rods to reduce the sodium void reactivity coefficient

Energy Technology Data Exchange (ETDEWEB)

Won, Jong Hyuck; Cho, Nam Zin, E-mail: nzcho@kaist.ac.kr; Park, Hae Min; Jeong, Yong Hoon, E-mail: jeongyh@kaist.ac.kr

2014-12-15

Highlights: • The graphite rod-inserted SFR fuel assembly is proposed to achieve low sodium void reactivity. • The neutronics/thermal-hydraulics analyses are performed for the proposed SFR cores. • The sodium void reactivity is improved about 960–1030 pcm compared to reference design. - Abstract: The concept of a graphite-moderating rod-inserted sodium-cooled fast reactor (SFR) fuel assembly is proposed in this study to achieve a low sodium void reactivity coefficient. Using this concept, two types of SFR cores are analyzed; the proposed SFR type 1 core has new SFR fuel assemblies at the inner/mid core regions while the proposed SFR type 2 core has a B{sub 4}C absorber sandwich in the middle of the active core region as well as new SFR fuel assemblies at the inner/mid core regions. For the proposed SFR core designs, neutronics and thermal-hydraulic analyses are performed using the DIF3D, REBUS3, and the MATRA-LMR codes. In the neutronics analysis, the sodium void reactivity coefficient is obtained in various void situations. The two types of proposed core designs reduce the sodium void reactivity coefficient by about 960–1030 pcm compared to the reference design. However, the TRU enrichment for the proposed SFR core designs is increased. In the thermal hydraulic analysis, the temperature distributions are calculated for the two types of proposed core designs and the mass flow rate is optimized to satisfy the design constraints for the highest power generating assembly. The results of this study indicate that the proposed SFR assembly design concept, which adopts graphite-moderating rods which are inserted into the fuel assembly, can feasibly minimize the sodium void reactivity coefficient. Single TRU enrichment and an identical fuel slug diameter throughout the SFR core are also achieved because the radial power peak can be flattened by varying the number of moderating rods in each core region.

Effects of fruit thinning, covering of the fruit truss and CO2 enrichment on radial fruit cracking in tomato [Lycopersicon esculentum] production under rain shelter in cool uplands

International Nuclear Information System (INIS)

Suzuki, T.; Nomura, Y.; Shimazu, T.; Tanaka, I.

2009-01-01

Radial fruit cracking (RFC) can contribute to serious economic losses in tomato production under rain shelter in cool uplands. In order to investigate the effects of translocation and distribution of photosynthate to the fruits during the occurrence of RFC, tomato plants were grown under treatments with fruit thinning and CO2 enrichment, which regulate the strength of sink and source, and treatments with covering of the fruit truss, which decreases solar radiation incident on the fruit surface. The occurrence of RFC was increased by fruit thinning and CO2 enrichment, and decreased by covering of fruit truss. Time course of the percentage of RFC to total harvest showed a remarkable rise toward the end of August and toward the end of October in 2004, when harvested fruit weight was increasing. These finding suggest that RFC is attributed to excessive enlargement of the fruit by promotion of translocation and distribution of photosynthate from leaves (source) to fruits (sink) and the solar radiation incident on the fruits. In addition, the relation between RFC and the generation of a cork layer is considered

Start-up fuel and power flattening of sodium-cooled candle core

International Nuclear Information System (INIS)

Takaki, Naoyuki; Sagawa, Yu; Umino, Akitake; Sekimoto, Hiroshi

2013-01-01

The hard neutron spectrum and unique power shape of CANDLE enable its distinctive performances such as achieving high burnup more than 30% and exempting necessity of both enrichment and reprocessing. On the other hand, they also cause several challenging problems. One is how the initial fuel can be prepared to start up the first CANDLE reactor because the equilibrium fuel composition that enables stable CANDLE burning is complex both in axial and radial directions. Another prominent problem is high radial power peaking factor that worsens averaged burnup, namely resource utilization factor in once-through mode and shorten the life time of structure materials. The purposes of this study are to solve these two problems. Several ideas for core configurations and startup fuel using single enrichment uranium and iron as a substitute of fission products are studied. As a result, it is found that low enriched uranium is applicable to ignite the core but all concepts examined here exceeded heat limits. Adjustment in enrichment and height of active and burnt zone is opened for future work. Sodium duct assemblies and thorium fuel assemblies loaded in the center region are studied as measures to reduce radial power peaking factor. Replacing 37 fuels by thorium fuel assemblies in the zeroth to third row provides well-balanced performance with flattened radial power distribution. The CANDLE core loaded with natural uranium in the outer and thorium in the center region achieved 35.6% of averaged burnup and 7.0 years of cladding life time owing to mitigated local fast neutron irradiation at the center. Using thorium with natural or depleted uranium in CANDLE reactor is also beneficial to diversifying fission resource and extending available term of fission energy without expansion of needs for enrichment and reprocessing

Preliminary study on the feasibility of ductless fuel assembly for fast reactors

International Nuclear Information System (INIS)

Shibahara, Itaru; Enokido, Yuji

1988-01-01

Preliminary study on the feasibility of ductless fuel assembly for fast reactors has been conducted. The primary concern is with forecasting the thermal hydraulic characteristics and the heat removal efficiency from the core. The thermal hydraulic analysis revealed the coolant mixing in the core at steady state operating condition was not intensive and the coolant temperature increase was almost proportional to the power of each assembly. The hot spot analysis of the ductless core indicated that the hottest temperature in the core could be comparable with the temperature of the conventional ducted core, even in case the radial power flattening was not actively pursued but with adopting ducted radial blanket assemblies. Under off-normal conditions, the ductless core had improved heat removal capability which was caused by inter-assembly coolant flow. The study has indicated the feasibility of the ductless fuel assembly for fast reactors. The experiments to demonstrate the feasibility will be the next key process for the development. (author)

First result from x-ray pulse height analyzer with radial scanning system for LHD

Science.gov (United States)

Muto, Sadatsugu; Morita, Shigeru

2001-01-01

Radial profiles of x-ray spectrum have been successfully obtained using an assembly of x-ray pulse height analyzer in large helical device. The observed profile is obtained from plasma heated by ICRF and neutral beam injection (NBI). As a detector, Si(Li) semiconductor is used with a histogramming memory and analog-to-digital converter (ADC) basically working at high counting rate up to 500 kcps. In routine operation a count rate of 62 kcps has been normally obtained with energy resolution better than 400 eV at iron Kα line. The assembly is equipped with four detectors and a radial scanning system which modulates sight lines of the detectors in major radius direction. The profiles of electron temperature and the intensity of metallic impurities have been obtained with a spatial resolution of a few centimeters. Measured electron temperature is in good agreement with that from Thomson scattering. The system is applicable to steady-state discharge. The design philosophy of the assembly and recent results on the performance tests are also presented.

A Preliminary Design Study of Ultra-Long-Life SFR Cores having Heterogeneous Fuel Assemblies

Energy Technology Data Exchange (ETDEWEB)

Jung, GeonHee; You, WuSeung; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of)

2016-10-15

The PWR and CANDU reactors have provided electricity for several decades in our country but they have produced lots of spent fuels and so the safe and efficient disposal of these spent fuels is one of the main issues in nuclear industry. This type ultra-long-life cores are quite efficient in terms of the amount of spent fuel generation per electricity production and they can be used as an interim storage for PWR or CANDU spent fuel over several tens of years if they use the PWR or CANDU spent fuel as the initial fuel. Typically, the previous works have considered radially homogeneous fuel assemblies in which only blanket or driver fuel rods are employed and they considered axially or radially heterogeneous core configurations with the radially homogeneous fuel assemblies. These core configurations result in the propagation of the power distribution which can lead to the significant temperature changes for each fuel assembly over the time. In this work, the radially heterogeneous fuel assemblies are employed in new ultra-long-life SFR (Sodium-cooled Fast Reactor) cores to minimize the propagation of power distribution by allowing the power propagation in the fuel assemblies. In this work, new small ultra-long life SFR cores were designed with heterogeneous fuel assemblies having both blanket and driver fuel rods to minimize the propagation of power distribution over the core by allowing power propagation from driver rods to blanket rods in fuel assemblies. In particular, high fidelity depletion calculation coupled with heterogeneous Monte Carlo neutron transport calculation was performed to assess the neutronic feasibility of the ultralong life cores. The results of the analysis showed that the candidate core has the cycle length of 77 EFPYs, a small burnup reactivity swing of 1590 pcm and acceptably small SVRs both at BOC and EOC.

ZPPR-20 phase D : a cylindrical assembly of polyethylene moderated U metal reflected by beryllium oxide and polyethylene.

Energy Technology Data Exchange (ETDEWEB)

Lell, R.; Grimm, K.; McKnight, R.; Shaefer, R.; Nuclear Engineering Division; INL

2006-09-30

The Zero Power Physics Reactor (ZPPR) fast critical facility was built at the Argonne National Laboratory-West (ANL-W) site in Idaho in 1969 to obtain neutron physics information necessary for the design of fast breeder reactors. The ZPPR-20D Benchmark Assembly was part of a series of cores built in Assembly 20 (References 1 through 3) of the ZPPR facility to provide data for developing a nuclear power source for space applications (SP-100). The assemblies were beryllium oxide reflected and had core fuel compositions containing enriched uranium fuel, niobium and rhenium. ZPPR-20 Phase C (HEU-MET-FAST-075) was built as the reference flight configuration. Two other configurations, Phases D and E, simulated accident scenarios. Phase D modeled the water immersion scenario during a launch accident, and Phase E (SUB-HEU-MET-FAST-001) modeled the earth burial scenario during a launch accident. Two configurations were recorded for the simulated water immersion accident scenario (Phase D); the critical configuration, documented here, and the subcritical configuration (SUB-HEU-MET-MIXED-001). Experiments in Assembly 20 Phases 20A through 20F were performed in 1988. The reference water immersion configuration for the ZPPR-20D assembly was obtained as reactor loading 129 on October 7, 1988 with a fissile mass of 167.477 kg and a reactivity of -4.626 {+-} 0.044{cents} (k {approx} 0.9997). The SP-100 core was to be constructed of highly enriched uranium nitride, niobium, rhenium and depleted lithium. The core design called for two enrichment zones with niobium-1% zirconium alloy fuel cladding and core structure. Rhenium was to be used as a fuel pin liner to provide shut down in the event of water immersion and flooding. The core coolant was to be depleted lithium metal ({sup 7}Li). The core was to be surrounded radially with a niobium reactor vessel and bypass which would carry the lithium coolant to the forward inlet plenum. Immediately inside the reactor vessel was a rhenium

Evaluation of efficiency of axial profiling in WWER-440 fuel assemblies

International Nuclear Information System (INIS)

Ananjev, Yu. A.; Kurakin, K. Yu.; Artemov, V.G.; Ivanov, A.S.

2005-01-01

The present report deals with consideration of fuel enrichment axial profiling in WWER-440 assemblies. The study is performed on improving the effectiveness of fuel utilization using the example of implementing the axial profiling in the assemblies of the second generation. For simulation of fuel loadings the computer code package SAPFIR 9 5 and RC is used that allows for correct consideration of specific features of assemblies design changes. The methodical approach to assessment of effectiveness of implementing the axial profiling is considered with the use of capabilities of the mentioned code package. In conclusion the recommendations are given on using the fuel enrichment axial profiling in WWER-440 assemblies (Authors)

A conceptual design of assembly strategy and dedicated tools for assembly of 40o sector

International Nuclear Information System (INIS)

Park, H.K.; Nam, K.O.; Kim, D.J.; Ahn, H.J.; Lee, J.H.; Im, K.; Shaw, R.

2010-01-01

The International Thermanuclear Experimental Reactor (ITER) tokamak device is composed of 9 vacuum vessel (VV)/toroidal field coils (TFCs)/vacuum vessel thermal shields (VVTS) 40 o sectors. Each VV/TFCs/VVTS 40 o sector is made up of one 40 o VV, two 20 o TFCs and associated VVTS segments. The 40 o sectors are sub-assembled at assembly hall respectively and then nine 40 o sectors sub-assembled at assembly hall are finally assembled at tokamak in-pit hall. The assembly strategy and tools for the 40 o sector sub-assembly and final assembly should be developed to satisfy the basic assembly requirements of the ITER tokamak device. Accordingly, the purpose-built assembly tools should be designed and manufactured considering assembly plan, available space, cost, safety, easy operation, efficient maintenance, and so on. The 40 o sector assembly tools are classified into 2 groups. One group is the sub-assembly tools including upending tool, lifting tool, sub-assembly tool, VV supports and bracing tools used at assembly hall and the other group is the in-pit assembly tools including lifting tool, central column, radial beams and their supports. This paper describes the current status of the assembly strategy and major tools for the VV/TFCs/VVTS 40 o sector assembly at in-pit hall and assembly hall. The conceptual design of the major assembly tools and assembly process at assembly hall and tokamak in-pit hall are presented also.

Apparatus for lifting spent fuel assembly

International Nuclear Information System (INIS)

Hirasawa, Yoshinari; Sato, Isao; Yoneda, Yoshiyuki.

1976-01-01

Object: To increase the efficiency of cooling of a used fuel assembly being moved within a guide tube in the axial direction thereof by directly cooling the assembly with cooling gas fed into the guide tube, thus facilitating the handling of the spent fuel assembly. Structure: An end of a lock portion is inserted into the top portion of a spent fuel assembly, the assembly being hooked on the lock portion. The lock portion is provided on its outer periphery with a seal member and a centering member and at its tip with a pawl capable of being projected and retracted in the radial direction. Thus, when the lock portion is moved along the guide tube, the used fuel assembly can be moved along the guide tube by maintaining the concentric relation thereto. Meanwhile, when cooling gas is fed into the guide tube, it is blown into the used fuel assembly to directly cool the same. Thus, the cooling efficiency can be increased. (Moriyama, M.)

HEAVY-ELEMENT ENRICHMENT OF A JUPITER-MASS PROTOPLANET AS A FUNCTION OF ORBITAL LOCATION

International Nuclear Information System (INIS)

Helled, R.; Schubert, G.

2009-01-01

One possible mechanism for giant planet formation is disk instability in which the planet is formed as a result of gravitational instability in the protoplanetary disk surrounding the young star. The final composition and core mass of the planet will depend on the planet's mass, environment, and the planetesimal accretion efficiency. We calculate heavy-element enrichment in a Jupiter-mass protoplanet formed by disk instability at various radial distances from the star, considering different disk masses and surface density distributions. Although the available mass for accretion increases with radial distance (a) for disk solid surface density (σ) functions σ = σ 0 a -α with α 5 years of planetary evolution, when the planet is extended and before gap opening and type II migration take place. The accreted mass is calculated for disk masses of 0.01, 0.05, and 0.1 M sun with α = 1/2, 1, and 3/2. We show that a Jupiter-mass protoplanet can accrete 1-110 M + of heavy elements, depending on the disk properties. Due to the limitation on the accretion timescale, our results provide lower bounds on heavy-element enrichment. Our results can explain the large variation in heavy-element enrichment found in extrasolar giant planets. Since higher disk surface density is found to lead to larger heavy-element enrichment, our model results are consistent with the correlation between heavy-element enrichment and stellar metallicity. Our calculations also suggest that Jupiter could have formed at a larger radial distance than its current location while still accreting the mass of heavy elements predicted by interior models. We conclude that in the disk instability model the final composition of a giant planet is strongly determined by its formation environment. The heavy-element abundance of a giant planet does not discriminate between its origin by either disk instability or core accretion.

Tuning porosity and radial mechanical properties of DNA origami nanotubes via crossover design

Science.gov (United States)

Ma, Zhipeng; Kawai, Kentaro; Hirai, Yoshikazu; Tsuchiya, Toshiyuki; Tabata, Osamu

2017-06-01

DNA origami nanotubes are utilized as structural platforms for the fabrication of various micro/nanosystems for drug delivery, optical or biological sensing, and even nanoscale robots. Their radial structural and mechanical properties, which play a crucial role in the effective use of micro/nanosystems, have not been fully studied. In particular, the effects of crossovers, which are basic structures for rationally assembling double-stranded DNA (dsDNA) helices into a nanotube configuration, have not yet been characterized experimentally. To investigate the effects of crossovers on the porosity and the radial mechanical properties of DNA origami nanotubes, we fabricated a DNA origami nanotube with varied crossover designs along the nanotube axis. The radial geometry of the DNA origami nanotube is experimentally characterized by both atomic force microscopy (AFM) and electron cryomicroscopy (cryo-EM). Moreover, the radial mechanical properties of the DNA origami nanotube including the radial modulus are directly measured by force-distance-based AFM. These measurements reveal that the porosity and the radial modulus of DNA origami nanotubes can be tuned by adjusting the crossover design, which enables the optimal design and construction of DNA origami nanostructures for various applications.

Fuel assembly and reactor core

International Nuclear Information System (INIS)

Yuchi, Yoko; Aoyama, Motoo; Haikawa, Katsumasa; Yamanaka, Akihiro; Koyama, Jun-ichi.

1996-01-01

In a fuel assembly of a BWR type reactor, a region substantially containing burnable poison is divided into an upper region and a lower region having different average concentrations of burnable poison along a transverse cross section perpendicular to the axial direction. The ratio of burnable poison contents of both regions is determined to not more than 80%, and the average concentration of the burnable poison in the lower region is determined to not less than 9% by weight. An infinite multiplication factor at an initial stage of the burning of the fuel assembly is controlled effectively by the burnable poisons. Namely, the ratio of the axial power can be controlled by the distribution of the enrichment degree of uranium fuels and the distribution of the burnable poison concentration in the axial direction. Since the average enrichment degree of the reactor core has to be increased in order to provide an initially loaded reactor core at high burnup degree. Distortion of the power distribution in the axial direction of the reactor core to which fuel assemblies at high enrichment degree are loaded is flattened to improve thermal margin, to extend continuous operation period and increase a burnup degree upon take-out thereby improving fuel economy without worsening the reactor core characteristics of the initially loaded reactor core. (N.H.)

Experimental validation of radial reconstructed pin-power distributions in full-scale BWR fuel assemblies with and without control blade

Energy Technology Data Exchange (ETDEWEB)

Giust, Flavio, E-mail: flavio.giust@axpo.c [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne, CH-1015 Lausanne (Switzerland); Axpo Kernenergie AG, Parkstrasse 23, CH-5401 Baden (Switzerland); Grimm, Peter [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Chawla, Rakesh [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne, CH-1015 Lausanne (Switzerland)

2010-12-15

Total fission rate measurements have been performed on full-size BWR fuel assemblies of type SVEA-96+ in the zero power reactor PROTEUS at the Paul Scherrer Institute. This paper presents comparisons of reconstructed 2D pin fission rates from nodal diffusion calculations to the experimental results in two configurations: one 'regular' (I-1A) and the other 'controlled' (I-2A). Both configurations consist of an array of 3 x 3 SVEA-96+ fuel assemblies moderated with light water at 20 {sup o}C. In configuration I-2A, an L-shaped hafnium control blade (half of a real cruciform blade) is inserted adjacent to the north-west corner of the central fuel assembly. To minimise the impact of the surroundings, all measurements were done in fuel pins belonging to the central assembly. The 3 x 3 experimental configuration (test zone) was modelled using the core monitoring and design tools that are applied at the Leibstadt Nuclear Power Plant (KKL). These are the 2D transport code HELIOS, used for the cross-section generation, and the 3D, 2-group nodal diffusion code PRESTO-2. The exterior is represented, in the axial and radial directions, by 2-group partial current ratios (PCRs) calculated at the test zone boundary using a 3D Monte Carlo (MCNPX) model of the whole PROTEUS reactor. Sensitivity cases are analysed to show the impact of changes in the 2D lattice modelling on the calculated fission rate distribution and reactivity. Further, the effects of variations in the test zone boundary PCRs and their behaviour in energy are investigated. For the test zone configuration without control blade, the pin-power reconstruction methodology delivers the same level of accuracy as the 2D transport calculations. On the other hand, larger deviations that are inherent to the use of reflected geometry in the lattice calculations are observed for the configuration with the control blade inserted. In the basic (reference) simulation cases, the calculated-to-experimental (C

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-03-19

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

Nuclear fuel assembly

International Nuclear Information System (INIS)

Ueda, Tomihiro.

1970-01-01

The present invention relates to fuel assemblies employing wire wrap spacers for retaining uniform spatial distribution between fuel elements. Clad fuel elements are helically wound in the oxial direction with a wave-formed wire strand. The strand is therefore provided with spring action which permits the fuel elements to expand freely in the axial and radial directions so as to retain proper spacing and reduce stresses due to thermal deformation. (Ownes, K.J.)

LEU WWR-M2 fuel assemblies burnable test

International Nuclear Information System (INIS)

Kirsanov, G.A.; Konoplev, K.A.; Pikulik, R.G.; Sajkov, Yu. P.; Tchmshkyan, D.V.; Tedoradze, L.V.; Zakharov, A.S.

2000-01-01

The results of in-pile irradiation tests of LEU WWR-M2 fuel assemblies with reduced enrichment of fuel are submitted in the report. The tests are made according to the Russian Program on Reduced Enrichment for Research and Test Reactors (RERTR). United States Department of Energy and the Ministry of Atomic Energy of Russian Federation jointly fund this Program. The irradiation tests of 5 WWR-M2 experimental assemblies are carried out at WWR-M reactor of the Petersburg Nuclear Physics Institute (PNPI). The information on assembly design and technique of irradiation tests is presented. In the irradiation tests the integrity of fuel assemblies is periodically measured. The report presents the data for the integrity maintained during the burnup of 5 fuel assemblies up to 45%. These results demonstrate the high reliability of the experimental fuel assemblies within the guaranteed burnup limits specified by the manufacturer. The tests are still in progress; it is planned to test and analyze the change in integrity for burnup of up to 70% - 75% or more. LEU WWR-M2 fuel assemblies are to be offered for export by their Novosibirsk manufacturer. Currently, HEU WWR-M2 fuel assemblies are used in Hungary, Ukraine and Vietnam. LEU WWR-M2 fuel assemblies were designed as a possible replacement for the HEU WWR-M2 fuel assemblies in those countries, but their use can be extended to other research reactors. (author)

Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

Science.gov (United States)

Favalli, A.; Vo, D.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S. J.; Trellue, H.; Vaccaro, S.

2016-06-01

The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/137Cs, 134Cs/137Cs, 106Ru/137Cs, and 144Ce/137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity's behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

Shaft seal assembly and method

Science.gov (United States)

Keba, John E. (Inventor)

2007-01-01

A pressure-actuated shaft seal assembly and associated method for controlling the flow of fluid adjacent a rotatable shaft are provided. The seal assembly includes one or more seal members that can be adjusted between open and closed positions, for example, according to the rotational speed of the shaft. For example, the seal member can be configured to be adjusted according to a radial pressure differential in a fluid that varies with the rotational speed of the shaft. In addition, in the closed position, each seal member can contact a rotatable member connected to the shaft to form a seal with the rotatable member and prevent fluid from flowing through the assembly. Thus, the seal can be closed at low speeds of operation and opened at high speeds of operation, thereby reducing the heat and wear in the seal assembly while maintaining a sufficient seal during all speeds of operation.

The ARCS radial collimator

International Nuclear Information System (INIS)

Stone, M.B.; Abernathy, D.L.; Niedziela, J.L.; Overbay, M.A.

2015-01-01

We have designed, installed, and commissioned a scattered beam radial collimator for use at the ARCS Wide Angular Range Chopper Spectrometer at the Spallation Neutron Source. The collimator has been designed to work effectively for thermal and epithermal neutrons and with a range of sample environments. Other design considerations include the accommodation of working within a high vacuum environment and having the ability to quickly install and remove the collimator from the scattered beam. The collimator is composed of collimating blades (or septa). The septa are 12 micron thick Kapton foils coated on each side with 39 microns of enriched boron carbide ( 10 B 4 C with 10 B > 96%) in an ultra-high vacuum compatible binder. The collimator blades represent an additional 22 m 2 of surface area. In the article we present collimator's design and performance and methodologies for its effective use

Composite airfoil assembly

Science.gov (United States)

Garcia-Crespo, Andres Jose

2015-03-03

A composite blade assembly for mounting on a turbine wheel includes a ceramic airfoil and an airfoil platform. The ceramic airfoil is formed with an airfoil portion, a blade shank portion and a blade dovetail tang. The metal platform includes a platform shank and a radially inner platform dovetail. The ceramic airfoil is captured within the metal platform, such that in use, the ceramic airfoil is held within the turbine wheel independent of the metal platform.

Bioavailability, rheology and sensory evaluation of fat-free yogurt enriched with VD3 encapsulated in re-assembled casein micelles.

Science.gov (United States)

Levinson, Yonatan; Ish-Shalom, Sophia; Segal, Elena; Livney, Yoav D

2016-03-01

Vitamin D3 (VD3) deficiency is a global problem. Better ways are needed to enrich foods with this important nutraceutical. VD3 is fat-soluble, hence requiring a suitable vehicle for enriching nonfat foods. Our objectives were to assess the bioavailability of VD3, from fat-free yogurt, in re-assembled casein micelles (rCMs) compared to that in polysorbate-80 (PS80/Tween-80) a commonly used synthetic emulsifier, and to assess and compare their rheology and palatability. We enriched fat-free yogurt with VD3 loaded into either rCM (VD3-rCMs) or PS80 (VD3-PS80). In vivo VD3 bioavailability was evaluated by a large randomized, double blind, placebo-controlled clinical trial, measuring serum 25(OH)D increase in subjects who consumed fat-free yogurt with 50,000 IU of either VD3-rCM, VD3-PS80, or VD3-free placebo yogurt. Both VD3-rCM and VD3-PS80 increased the serum 25(OH)D levels by ∼8 ng ml(-1) and no significant differences in mean 25(OH)D levels were observed, evidencing the fact that VD3 bioavailability in rCM was as high as that in the synthetic emulsifier. VD3-rCM yogurt had a higher viscosity than VD3-PS80 yogurt. In sensory evaluations, panelists were able to discern between VD3-rCM and VD3-PS80 yogurt, and showed a dislike for PS80 yogurt, compared to rCM or the unenriched control. These results complement our past results showing higher protection against thermal treatment, UV irradiation, and deterioration during shelf life, conferred to hydrophobic nutraceuticals by rCM compared to that by the synthetic surfactant or to the unprotected bioactive, in showing the advantageous use of rCM over the synthetic emulsifier as a delivery system for the enrichment of food with VD3 and other hydrophobic nutraceuticals.

Reactor physics studies in the GCFR phase-II critical assembly

International Nuclear Information System (INIS)

Pond, R.B.

1976-09-01

The reactor physics studies performed in the gas cooled fast reactor (GCFR) mockup on ZPR-9 are covered. This critical assembly, designated Phase II in the GCFR program, had a single zone PuO 2 -UO 2 core composition and UO 2 radial and axial blankets. The assembly was built both with and without radial and axial stainless steel reflectors. The program included the following measurements: small-sample reactivity worths of reactor constituent materials (including helium); 238 U Doppler effect; uranium and plutonium reaction rate distributions; thorium, uranium, and plutonium α and reactor kinetics. Analysis of the measurements used ENDF/B-IV nuclear data; anisotropic diffusion coefficients were used to account for neutron streaming effects. Comparison of measurements and calculations to GCFR Phase I are also made

Hyb-Seq: combining target enrichment and genome skimming for plant phylogenomics

Science.gov (United States)

Kevin Weitemier; Shannon C.K. Straub; Richard C. Cronn; Mark Fishbein; Roswitha Schmickl; Angela McDonnell; Aaron. Liston

2014-01-01

• Premise of the study: Hyb-Seq, the combination of target enrichment and genome skimming, allows simultaneous data collection for low-copy nuclear genes and high-copy genomic targets for plant systematics and evolution studies. • Methods and Results: Genome and transcriptome assemblies for milkweed ( Asclepias syriaca ) were used to design enrichment probes for 3385...

Refueling the RPI reactor facility with low-enrichment fuel

International Nuclear Information System (INIS)

Harris, D.R.; Rodriguez-Vera, F.; Wicks, F.E.

1985-01-01

The RPI Critical Facility has operated since 1963 with a core of thin, highly enriched fuel plates in twenty-five fuel assembly boxes. A program is underway to refuel the reactor with 4.81 w/o enriched SPERT (F-1) fuel rods. Use of these fuel rods will upgrade the capabilities of the reactor and will eliminate a security risk. Adequate quantities of SPERT (F-1) fuel rods are available, and their use will result in a great cost saving relative to manufacturing new low-enrichment fuel plates. The SPERT fuel rods are 19 inches longer than are the present fuel plates, so a modified core support structure is required. It is planned to support and position the SPERT fuel pins by upper and lower lattice plates, thus avoiding the considerable cost of new fuel assembly boxes. The lattice plates will be secured to the existing top and bottom plates. The design permits the fabrication and use of other lattice plates for critical experiment research programs in support of long-lived full development for power reactors. (author)

Structural analysis of ITER sub-assembly tools

International Nuclear Information System (INIS)

Nam, K.O.; Park, H.K.; Kim, D.J.; Ahn, H.J.; Lee, J.H.; Kim, K.K.; Im, K.; Shaw, R.

2011-01-01

The ITER Tokamak assembly tools are purpose-built assembly tools to complete the ITER Tokamak machine which includes the cryostat and the components contained therein. The sector sub-assembly tools descried in this paper are main assembly tools to assemble vacuum vessel, thermal shield and toroidal filed coils into a complete 40 o sector. The 40 o sector sub-assembly tools are composed of sector sub-assembly tool, including radial beam, vacuum vessel supports and mid-plane brace tools. These tools shall have sufficient strength to transport and handle heavy weight of the ITER Tokamak machine reached several hundred tons. Therefore these tools should be designed and analyzed to confirm both the strength and structural stability even in the case of conservative assumptions. To verify structural stabilities of the sector sub-assembly tools in terms of strength and deflection, ANSYS code was used for linear static analysis. The results of the analysis show that these tools are designed with sufficient strength and stiffness. The conceptual designs of these tools are briefly described in this paper also.

Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

Energy Technology Data Exchange (ETDEWEB)

Favalli, A., E-mail: afavalli@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM (United States); Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Schwalbach, P. [European Commission, DG Energy, Euratom Safeguards Luxemburg, Luxemburg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Tobin, S.J.; Trellue, H. [Los Alamos National Laboratory, Los Alamos, NM (United States); Vaccaro, S. [European Commission, DG Energy, Euratom Safeguards Luxemburg, Luxemburg (Luxembourg)

2016-06-01

The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute {sup 137}Cs count rate and the {sup 154}Eu/{sup 137}Cs, {sup 134}Cs/{sup 137}Cs, {sup 106}Ru/{sup 137}Cs, and {sup 144}Ce/{sup 137}Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

Fuel assemblies

International Nuclear Information System (INIS)

Yoshioka, Ritsuo.

1983-01-01

Purpose: To improve the operation performance of a BWR type reactor by improving the distribution of the uranium enrichment and the incorporation amount of burnable poisons in fuel assemblies. Constitution: The average enrichment of uranium 235 is increased in the upper portion as compared with that in the lower portion, while the incorporation amount of burnable poisons is increased in an upper portion as compared with that in the lower portion. The difference in the incorporation amount of the burnable poisons between the upper and lower portions is attained by charging two kinds of fuel rods; the ones incorporated with the burnable poisons over the entire length and the others incorporated with the burnable poisons only in the upper portions. (Seki, T.)

Extrapolated experimental critical parameters of unreflected and steel-reflected massive enriched uranium metal spherical and hemispherical assemblies

International Nuclear Information System (INIS)

Rothe, R.E.

1997-12-01

Sixty-nine critical configurations of up to 186 kg of uranium are reported from very early experiments (1960s) performed at the Rocky Flats Critical Mass Laboratory near Denver, Colorado. Enriched (93%) uranium metal spherical and hemispherical configurations were studied. All were thick-walled shells except for two solid hemispheres. Experiments were essentially unreflected; or they included central and/or external regions of mild steel. No liquids were involved. Critical parameters are derived from extrapolations beyond subcritical data. Extrapolations, rather than more precise interpolations between slightly supercritical and slightly subcritical configurations, were necessary because experiments involved manually assembled configurations. Many extrapolations were quite long; but the general lack of curvature in the subcritical region lends credibility to their validity. In addition to delayed critical parameters, a procedure is offered which might permit the determination of prompt critical parameters as well for the same cases. This conjectured procedure is not based on any strong physical arguments

Radial nerve dysfunction

Science.gov (United States)

Neuropathy - radial nerve; Radial nerve palsy; Mononeuropathy ... Damage to one nerve group, such as the radial nerve, is called mononeuropathy . Mononeuropathy means there is damage to a single nerve. Both ...

Physics analyses of an accelerator-driven sub-critical assembly

Science.gov (United States)

Naberezhnev, Dmitry G.; Gohar, Yousry; Bailey, James; Belch, Henry

2006-06-01

Physics analyses have been performed for an accelerator-driven sub-critical assembly as a part of the Argonne National Laboratory activity in preparation for a joint conceptual design with the Kharkov Institute of Physics and Technology (KIPT) of Ukraine. KIPT has a plan to construct an accelerator-driven sub-critical assembly targeted towards the medical isotope production and the support of the Ukraine nuclear industry. The external neutron source is produced either through photonuclear reactions in tungsten or uranium targets, or deuteron reactions in a beryllium target. KIPT intends using the high-enriched uranium (HEU) for the fuel of the sub-critical assembly. The main objective of this paper is to study the possibility of utilizing low-enriched uranium (LEU) fuel instead of HEU fuel without penalizing the sub-critical assembly performance, in particular the neutron flux level. In the course of this activity, several studies have been carried out to investigate the main choices for the system's parameters. The external neutron source has been characterized and a pre-conceptual target design has been developed. Several sub-critical configurations with different fuel enrichments and densities have been considered. Based on our analysis, it was shown that the performance of the LEU fuel is comparable with that of the HEU fuel. The LEU fuel sub-critical assembly with 200-MeV electron energy and 100-kW electron beam power has an average total flux of ˜2.50×10 13 n/s cm 2 in the irradiation channels. The corresponding total facility power is ˜204 kW divided into 91 and 113 kW deposited in the target and sub-critical assemblies, respectively.

Insulation assembly for electric machine

Science.gov (United States)

Rhoads, Frederick W.; Titmuss, David F.; Parish, Harold; Campbell, John D.

2013-10-15

An insulation assembly is provided that includes a generally annularly-shaped main body and at least two spaced-apart fingers extending radially inwards from the main body. The spaced-apart fingers define a gap between the fingers. A slot liner may be inserted within the gap. The main body may include a plurality of circumferentially distributed segments. Each one of the plurality of segments may be operatively connected to another of the plurality of segments to form the continuous main body. The slot liner may be formed as a single extruded piece defining a plurality of cavities. A plurality of conductors (extendable from the stator assembly) may be axially inserted within a respective one of the plurality of cavities. The insulation assembly electrically isolates the conductors in the electric motor from the stator stack and from other conductors.

Molten core retention assembly

International Nuclear Information System (INIS)

Lampe, R.F.

1976-01-01

Molten fuel produced in a core overheating accident is caught by a molten core retention assembly consisting of a horizontal baffle plate having a plurality of openings therein, heat exchange tubes having flow holes near the top thereof mounted in the openings, and a cylindrical imperforate baffle attached to the plate and surrounding the tubes. The baffle assembly is supported from the core support plate of the reactor by a plurality of hanger rods which are welded to radial beams passing under the baffle plate and intermittently welded thereto. Preferably the upper end of the cylindrical baffle terminates in an outwardly facing lip to which are welded a plurality of bearings having slots therein adapted to accept the hanger rods

Criticality safety evaluation report for FFTF 42% fuel assemblies

International Nuclear Information System (INIS)

Richard, R.F.

1997-01-01

An FFTF tritium/isotope production mission will require a new fuel supply. The reference design core will use a mixed oxide fuel nominally enriched to 40 wt% Pu. This enrichment is significantly higher than that of the standard Driver Fuel Assemblies used in past operations. Consequently, criticality safety for handling and storage of this fuel must be addressed. The purpose of this document is to begin the process by determining the minimum critical number for these new fuel assemblies in water, sodium and air. This analysis is preliminary and further work can be done to refine the results reported here. Analysis was initially done using 45 wt 5 PuO. Additionally, a preliminary assessment is done concerning storage of these fuel assemblies in Interim Decay Storage (IDS), Fuel Storage Facility (FSF), and Core Component Containers/Interim Storage Casks (CCC/ISC)

Radial head button holing: a cause of irreducible anterior radial head dislocation

Energy Technology Data Exchange (ETDEWEB)

Shin, Su-Mi; Chai, Jee Won; You, Ja Yeon; Park, Jina [Seoul National University Seoul Metropolitan Government Boramae Medical Center, Department of Radiology, Seoul (Korea, Republic of); Bae, Kee Jeong [Seoul National University Seoul Metropolitan Government Boramae Medical Center, Department of Orthopedic Surgery, Seoul (Korea, Republic of)

2016-10-15

''Buttonholing'' of the radial head through the anterior joint capsule is a known cause of irreducible anterior radial head dislocation associated with Monteggia injuries in pediatric patients. To the best of our knowledge, no report has described an injury consisting of buttonholing of the radial head through the annular ligament and a simultaneous radial head fracture in an adolescent. In the present case, the radiographic findings were a radial head fracture with anterior dislocation and lack of the anterior fat pad sign. Magnetic resonance imaging (MRI) clearly demonstrated anterior dislocation of the fractured radial head through the torn annular ligament. The anterior joint capsule and proximal portion of the annular ligament were interposed between the radial head and capitellum, preventing closed reduction of the radial head. Familiarity with this condition and imaging findings will aid clinicians to make a proper diagnosis and fast decision to perform an open reduction. (orig.)

The proposed use of low enriched uranium fuel in the High Flux Australian Reactor (HIFAR)

International Nuclear Information System (INIS)

Vittorio, D.; Durance, G.

2002-01-01

The Australian Nuclear Science and Technology Organisation (ANSTO) operates the High Flux Australian Reactor (HIFAR). HIFAR commenced operation in the late 1950's with fuel elements containing uranium enriched to 93%. From that time the level of enrichment has gradually decreased to the current level of 60%. It is now proposed to further reduce the enrichment of HIFAR fuel to <20% by utilising LEU fuel assemblies manufactured by RISO National Laboratory, that were originally intended for use in the DR-3 reactor. Minor modifications have been made to the assemblies to adapt them for use in HIFAR. A detailed design review has been performed and initial safety analysis and reactor physics calculations are to be submitted to ARPANSA as part of a four-stage approval process. (author)

Criticality safety study of dry spent fuel cask loaded with increased enrichment fuel

International Nuclear Information System (INIS)

Bznuni, S.; Baghdasaryan, N.; Amirjanyan, A.

2013-01-01

Existing Dry Spent Fuel Casks (DSC) for transporting and storing of Armenian NPP fuel was licensed for WWER-440 fuel assemblies with 3.6% enrichment. Having in mind that ANPP introduced new fuel assemblies with increased enrichment (3.82 %) re-assessment of criticality safety analysis for DSC is required. Criticality safety analysis of DSC was performed by KENO-VI program using 238-GROUP ENDF/B-VII.0 LIBRARY (V7-238). Results of analysis showed that additional 8 borated racks for fuel assemblies should be included in the design of DSC. In addition feasibility study was performed to find out level of burnup-credit approach implementation to keep current design of DSC unchanged. Burnup-credit analysis was performed by STARBUCS program using axial burnup profiles from Armenian NPP neutronics analysis carried out by BIPR code. (authors)

Stability of radial and non-radial pulsation modes of massive ZAMS models

International Nuclear Information System (INIS)

Odell, A.P.; Pausenwein, A.; Weiss, W.W.; Hajek, A.

1987-01-01

The authors have computed non-adiabatic eigenvalues for radial and non-radial pulsation modes of star models between 80 and 120 M solar with composition of chi=0.70 and Z=0.02. The radial fundamental mode is unstable in models with mass greater than 95 M solar , but the first overtone mode is always stable. The non-radial modes are all stable for all models, but the iota=2 f-mode is the closest to being driven. The non-radial modes are progressively more stable with higher iota and with higher n (for both rho- and g-modes). Thus, their results indicate that radial pulsation limits the upper mass of a star

Influence of FA pin power minimisation on neutron-physical core characteristics

International Nuclear Information System (INIS)

Mikolas, P.

2005-01-01

The influence of optimisation of radial fuel enrichment profilation in fuel assembly of Gd-2 type on lowering its pin power non-uniformity and subsequent lowering of F dH in WWER-440 cores are described in this paper (Author)

Nickel container of highly-enriched uranium bodies and sodium

Science.gov (United States)

Zinn, Walter H.

1976-01-01

A fuel element comprises highly a enriched uranium bodies coated with a nonfissionable, corrosion resistant material. A plurality of these bodies are disposed in layers, with sodium filling the interstices therebetween. The entire assembly is enclosed in a fluid-tight container of nickel.

Nickel container of highly-enriched uranium bodies and sodium

International Nuclear Information System (INIS)

Zinn, W.H.

1976-01-01

A fuel element comprises highly enriched uranium bodies coated with a nonfissionable, corrosion resistant material. A plurality of these bodies are disposed in layers, with sodium filling the interstices therebetween. The entire assembly is enclosed in a fluid-tight container of nickel

Biomimetic fiber assembled gradient hydrogel to engineer glycosaminoglycan enriched and mineralized cartilage: An in vitro study.

Science.gov (United States)

Mohan, Neethu; Wilson, Jijo; Joseph, Dexy; Vaikkath, Dhanesh; Nair, Prabha D

2015-12-01

The study investigated the potential of electrospun fiber assembled hydrogel, with physical gradients of chondroitin sulfate (CS) and sol-gel-derived bioactive glass (BG), to engineer hyaline and mineralized cartilage in a single 3D system. Electrospun poly(caprolactone) (PCL) fibers incorporated with 0.1% w/w of CS (CSL) and 0.5% w/w of CS (CSH), 2.4% w/w of BG (BGL) and 12.5% w/w of BG (BGH) were fabricated. The CS showed a sustained release up to 3 days from CSL and 14 days from CSH fibers. Chondrocytes secreted hyaline like matrix with higher sulfated glycosaminoglycans (sGAG), collagen type II and aggrecan on CSL and CSH fibers. Mineralization was observed on BGL and BGH fibers when incubated in simulated body fluid for 14 days. Chondrocytes cultured on these fibers secreted a mineralized matrix that consisted of sGAG, hypertrophic proteins, collagen type X, and osteocalcin. The CS and BG incorporated PCL fiber mats were assembled in an agarose-gelatin hydrogel to generate a 3D hybrid scaffold. The signals in the fibers diffused and generated continuous opposing gradients of CS (chondrogenic signal) and BG (mineralization) in the hydrogel. The chondrocytes were encapsulated in hybrid scaffolds; live dead assay at 48 h showed viable cells. Cells maintained their phenotype and secreted specific extracellular matrix (ECM) in response to signals within the hydrogel. Continuous opposing gradients of sGAG enriched and mineralized ECM were observed surrounding each cell clusters on gradient hydrogel after 14 days of culture in response to the physical gradients of raw materials CS and BG. A construct with gradient mineralization might accelerate integration to subchondral bone during in vivo regeneration. © 2015 Wiley Periodicals, Inc.

The MARVEL assembly for neutron multiplication

Energy Technology Data Exchange (ETDEWEB)

David L. Chichester; Mathew T. Kinlaw

2013-10-01

A new multiplying test assembly is under development at Idaho National Laboratory to support research, validation, evaluation, and learning. The item is comprised of three stacked, highly-enriched uranium (HEU) cylinders, each 11.4 cm in diameter and having a combined height of up to 11.7 cm. The combined mass of all three cylinders is 20.3 kg of HEU. Calculations for the bare configuration of the assembly indicate a multiplication level of >3.5 (keff=0.72). Reflected configurations of the assembly, using either polyethylene or tungsten, are possible and have the capability of raising the assembly's multiplication level to greater than 10. This paper describes simulations performed to assess the assembly's multiplication level under different conditions and describes the resources available at INL to support the use of these materials. We also describe some preliminary calculations and test activities using the assembly to study neutron multiplication.

The MARVEL assembly for neutron multiplication.

Science.gov (United States)

Chichester, David L; Kinlaw, Mathew T

2013-10-01

A new multiplying test assembly is under development at Idaho National Laboratory to support research, validation, evaluation, and learning. The item is comprised of three stacked, highly-enriched uranium (HEU) cylinders, each 11.4 cm in diameter and having a combined height of up to 11.7 cm. The combined mass of all three cylinders is 20.3 kg of HEU. Calculations for the bare configuration of the assembly indicate a multiplication level of >3.5 (k(eff)=0.72). Reflected configurations of the assembly, using either polyethylene or tungsten, are possible and have the capability of raising the assembly's multiplication level to greater than 10. This paper describes simulations performed to assess the assembly's multiplication level under different conditions and describes the resources available at INL to support the use of these materials. We also describe some preliminary calculations and test activities using the assembly to study neutron multiplication. Copyright © 2013 Elsevier Ltd. All rights reserved.

Gag induces the coalescence of clustered lipid rafts and tetraspanin-enriched microdomains at HIV-1 assembly sites on the plasma membrane.

Science.gov (United States)

Hogue, Ian B; Grover, Jonathan R; Soheilian, Ferri; Nagashima, Kunio; Ono, Akira

2011-10-01

The HIV-1 structural protein Gag associates with two types of plasma membrane microdomains, lipid rafts and tetraspanin-enriched microdomains (TEMs), both of which have been proposed to be platforms for HIV-1 assembly. However, a variety of studies have demonstrated that lipid rafts and TEMs are distinct microdomains in the absence of HIV-1 infection. To measure the impact of Gag on microdomain behaviors, we took advantage of two assays: an antibody-mediated copatching assay and a Förster resonance energy transfer (FRET) assay that measures the clustering of microdomain markers in live cells without antibody-mediated patching. We found that lipid rafts and TEMs copatched and clustered to a greater extent in the presence of membrane-bound Gag in both assays, suggesting that Gag induces the coalescence of lipid rafts and TEMs. Substitutions in membrane binding motifs of Gag revealed that, while Gag membrane binding is necessary to induce coalescence of lipid rafts and TEMs, either acylation of Gag or binding of phosphatidylinositol-(4,5)-bisphosphate is sufficient. Finally, a Gag derivative that is defective in inducing membrane curvature appeared less able to induce lipid raft and TEM coalescence. A higher-resolution analysis of assembly sites by correlative fluorescence and scanning electron microscopy showed that coalescence of clustered lipid rafts and TEMs occurs predominantly at completed cell surface virus-like particles, whereas a transmembrane raft marker protein appeared to associate with punctate Gag fluorescence even in the absence of cell surface particles. Together, these results suggest that different membrane microdomain components are recruited in a stepwise manner during assembly.

Criticality calculations for a critical assembly, graphite moderate, using 20% enriched uranium

International Nuclear Information System (INIS)

Almeida Ferreira, A.C. de; Hukai, R.Y.

1975-01-01

The construction of a Zero Power Reactor (ZPR) at the Instituto de Energia Atomica in order to measure the neutron characteristics (parameters) of HTGR reactors is proposed. The necessary quantity fissile uranium for these measurements has been calculed. Criticality studies of graphite moderated critical assemblies containing thorium have been made and the critical mass of each of several typical commercial HTGR compositions has been calculated using computer codes HAMMER and CITATION. Assemblies investigated contained a central cylindrical core region, simulating a typical commercial HTGR composition, a uranium-graphite driver region and a outer pure graphite reflector region. It is concluded that a 10Kg inventory of fissile uranium will be required for a program of measurements utilizing each of the several calculated assemblies

Heterogeneous assembly for plutonium multi recycling in PWRs: the Corail concept

International Nuclear Information System (INIS)

Youinou, G.; Zaetta, A.; Vasile, A.; Delpech, M.; Rohart, M.; Guillet, J.L.

2001-01-01

The CORAIL assembly is a standard 17 x 17 PWR fuel assembly containing 180 UO 2 rods and 84 MOX rods located at the periphery to limit the hot-channel factor. After many recycling, the plutonium content stabilizes around 8% and the U 235 enrichment around 4.8% (for a 3 u 15000 MWd/t fuel cycle length). An all-CORAIL park would have a zero plutonium mass balance, and compared with an all-UO 2 park the gain in terms of Separating Work Units and natural uranium would be between 15% and 20%. Detailed calculations of a 1300 MWe PWR loaded with such assemblies show that its control would not require the use of enriched boron. Burnable poison is necessary to limit the hot-channel factor. (author)

A conceptual design and structural stabilities of in-pit assembly tools for the completion of final sector assembly at tokamak hall

International Nuclear Information System (INIS)

Nam, K.O.; Park, H.K.; Kim, D.J.; Ahn, H.J.; Kim, K.K.; Im, K.; Shaw, R.

2010-01-01

The final assembly of main components of the International Thermonuclear Experimental Reactor (ITER) tokamak, Vacuum Vessel (VV) and Toroidal Field Coils (TFCs), is achieved by the sequential assembly of the nine sub-assembled 40 o sectors in tokamak pit. Each sub-assembled 40 o sector is composed of one VV 40 o sector, two TFCs, and in-between Vacuum Vessel Thermal Shield (VVTS) segments. Sub-assembly is carried out in the assembly building and then the sub-assembled sectors are transferred into tokamak pit, in sequence, to complete sector assembly. The role of in-pit assembly tool is to support and align the sub-assembled sectors in tokamak pit. It also plays the role of reference datum during assembly until the completion of main components assembly. Korea Domestic Agency (KO DA) has developed the conceptual design of most ITER purpose-built assembly tools under the collaboration with the ITER Organization. Among the conceptual designs carried out, this paper describes the function, the structure, the selected material and the design results of the in-pit assembly tools comprising central column, radial beams and their supports, TF inner supports and in-pit working floor. The results of structural analysis using ANSYS for the various loading cases are given as well. The resultant stresses and deflections turned out to fall within the allowable ranges.

Uranium enrichment management review: summary of analysis

International Nuclear Information System (INIS)

1981-01-01

In May 1980, the Assistant Secretary for Resource Applications within the Department of Energy requested that a group of experienced business executives be assembled to review the operation, financing, and management of the uranium enrichment enterprise as a basis for advising the Secretary of Energy. After extensive investigation, analysis, and discussion, the review group presented its findings and recommendations in a report on December 2, 1980. The following pages contain background material on which that final report was based. This report is arranged in chapters that parallel those of the uranium enrichment management review final report - chapters that contain summaries of the review group's discussion and analyses in six areas: management of operations and construction; long-range planning; marketing of enrichment services; financial management; research and development; and general management. Further information, in-depth analysis, and discussion of suggested alternative management practices are provided in five appendices

Uranium enrichment management review: summary of analysis

Energy Technology Data Exchange (ETDEWEB)

1981-01-01

In May 1980, the Assistant Secretary for Resource Applications within the Department of Energy requested that a group of experienced business executives be assembled to review the operation, financing, and management of the uranium enrichment enterprise as a basis for advising the Secretary of Energy. After extensive investigation, analysis, and discussion, the review group presented its findings and recommendations in a report on December 2, 1980. The following pages contain background material on which that final report was based. This report is arranged in chapters that parallel those of the uranium enrichment management review final report - chapters that contain summaries of the review group's discussion and analyses in six areas: management of operations and construction; long-range planning; marketing of enrichment services; financial management; research and development; and general management. Further information, in-depth analysis, and discussion of suggested alternative management practices are provided in five appendices.

Optimization of seed-blanket type fuel assembly for reduced-moderation water reactor

Energy Technology Data Exchange (ETDEWEB)

Shelley, Afroza; Shimada, Shoichiro; Kugo, Teruhiko; Okubo, Tsutomu E-mail: okubo@hems.jaeri.go.jp; Iwamura, Takamichi

2003-10-01

Parametric studies have been performed for a PWR-type reduced-moderation water reactor (RMWR) with the seed-blanket type fuel assembles to achieve a high conversion ratio, negative void reactivity coefficient and a high burnup by using MOX fuel. From the viewpoint of reactor safety analysis, the fuel temperature coefficients were also studied. From the result of the burnup calculation, it has been seen that ratio of 40-50% of outer blanket in a seed-blanket assembly gives higher conversion ratio compared to the other combination of seed-blanket assembly. And the recommended number of (seed+blanket) layers is 20, in which the number of seed (S) layers is 15 (S15) and blanket (B) layers is 5 (B5). It was found that the conversion ratio of seed-blanket assembly decreases, when they are arranged looks like a flower shape (Hanagara). By the optimization of different parameters, S15B5 fuel assembly with the height of seed of 1000 mmx2, internal blanket of 150 mm and axial blanket of 400 mmx2 is recommended for a reactor of high conversion ratio. In this assembly, the gap of seed fuel rod is 1.0 mm and blanket fuel rod is 0.4 mm. In S15B5 assembly, the conversion ratio is 1.0 and the burnup is 38.18 GWd/t in (seed+internal blanket+outer blanket) region. However, the burnup is 57.45 GWd/t in (seed+internal blanket) region. The cycle length of the core is 16.46 effective full power in month (EFPM) by six batches and the enrichment of fissile Pu is 14.64 wt.%. The void coefficient is +21.82 pcm/%void, however, it is expected that the void coefficient will be negative if the radial neutron leakage is taken into account in the calculation. It is also possible to use S15B5 fuel assembly as a high burnup reactor 45 GWd/t in (seed+internal blanket+outer blanket) region, however, it is necessary to decrease the height of seed to 500 mmx2 to improve the void coefficient. In this reactor, the conversion ratio is 0.97 and void coefficient is +20.81 pcm/%void. The fuel temperature

Fuel assembly

International Nuclear Information System (INIS)

Ueda, Makoto; Ogiya, Shunsuke.

1989-01-01

For improving the economy of a BWR type reactor by making the operation cycle longer, the fuel enrichment degree has to be increased further. However, this makes the subcriticality shallower in the upper portion of the reactor core, to bring about a possibility that the reactor shutdown becomes impossible. In the present invention, a portion of fuel rod is constituted as partial length fuel rods (P-fuel rods) in which the entire stack length in the effective portion is made shorter by reducing the concentration of fissionable materials in the axial portion. A plurality of moderator rods are disposed at least on one diagonal line of a fuel assembly and P-fuel rods are arranged at a position put between the moderator rods. This makes it possible to reactor shutdown and makes the axial power distribution satisfactory even if the fuel enrichment degree is increased. (T.M.)

Paired replacement fuel assemblies for BWR-type reactor

International Nuclear Information System (INIS)

Oguchi, Kazushige.

1997-01-01

There are disposed a large-diameter water rod constituting a non-boiling region at a central portion and paired replacement fuel assemblies for two streams having the same average enrichment degree and different amount of burnable poisons. The paired replacement fuel assemblies comprise a first fuel assembly having a less amount of burnable poisons and a second fuel assembly having a larger amount of burnable poisons. A number of burnable poison-containing fuel rods in adjacent with the large diameter water rod is increased in the second fuel assembly than the first fuel assembly. Then, the poison of the paired replacement fuel assemblies for the BWR type reactor can be annihilated simultaneously at the final stage of the cycle. Accordingly, fuels for a BWR type reactor excellent in economical property and safety and facilitating the design of the replacement reactor core can be obtained. (N.H.)

Temperature and boron dependencies of buckling and radial reflector saving for VVER lattices

International Nuclear Information System (INIS)

Alvarez, C.

1990-01-01

The temperature and boron dependencies of buckling and radial reflectors savings are analyzed in this paper on the basis of the results from the calculations ZR-6M critical assembly. These dependencies are related to the physical behavior of temperature and boron reactivity coefficients for the cores of VVER-type critical facilities. As a byproduct, the parameter was also investigated and its dependence on water density was determined

Gag Induces the Coalescence of Clustered Lipid Rafts and Tetraspanin-Enriched Microdomains at HIV-1 Assembly Sites on the Plasma Membrane ▿

Science.gov (United States)

Hogue, Ian B.; Grover, Jonathan R.; Soheilian, Ferri; Nagashima, Kunio; Ono, Akira

2011-01-01

The HIV-1 structural protein Gag associates with two types of plasma membrane microdomains, lipid rafts and tetraspanin-enriched microdomains (TEMs), both of which have been proposed to be platforms for HIV-1 assembly. However, a variety of studies have demonstrated that lipid rafts and TEMs are distinct microdomains in the absence of HIV-1 infection. To measure the impact of Gag on microdomain behaviors, we took advantage of two assays: an antibody-mediated copatching assay and a Förster resonance energy transfer (FRET) assay that measures the clustering of microdomain markers in live cells without antibody-mediated patching. We found that lipid rafts and TEMs copatched and clustered to a greater extent in the presence of membrane-bound Gag in both assays, suggesting that Gag induces the coalescence of lipid rafts and TEMs. Substitutions in membrane binding motifs of Gag revealed that, while Gag membrane binding is necessary to induce coalescence of lipid rafts and TEMs, either acylation of Gag or binding of phosphatidylinositol-(4,5)-bisphosphate is sufficient. Finally, a Gag derivative that is defective in inducing membrane curvature appeared less able to induce lipid raft and TEM coalescence. A higher-resolution analysis of assembly sites by correlative fluorescence and scanning electron microscopy showed that coalescence of clustered lipid rafts and TEMs occurs predominately at completed cell surface virus-like particles, whereas a transmembrane raft marker protein appeared to associate with punctate Gag fluorescence even in the absence of cell surface particles. Together, these results suggest that different membrane microdomain components are recruited in a stepwise manner during assembly. PMID:21813604

MOX fuel assembly and reactor core

International Nuclear Information System (INIS)

Shimada, Hidemitsu; Koyama, Jun-ichi; Aoyama, Motoo

1998-01-01

The MOX fuel assembly of the present invention is of a c-lattice type loaded to a BWR type reactor. 74 MOX fuel rods filled with mixed oxides of uranium and plutonium and two water rods disposed to a space equal to that for 7 MOX fuel rods are arranged in 9 x 9 matrix. MOX fuel rods having the lowest enrichment degree are disposed to four corners of the 9 x 9 matrix. The enrichment degree means a ratio of the weight of fission products based on the total weight of fuels. Two MOX fuel rods having the same enrichment degree are arranged in each direction so as to be continuous from the MOX fuel rods at four corners in the direction of the same row and different column and same column and the different row. In addition, among the outermost circumferential portion of the 9 x 9 matrix, MOX fuel rods having a lower enrichment degree next to the MOX fuel rods having the lowest enrichment degree are arranged, each by three to a portion where MOX fuel rods having the lowest enrichment degree are not disposed. (I.N.)

Spontaneous Assembly of Exopolymers from Phytoplankton

Directory of Open Access Journals (Sweden)

Yong-Xue Ding

2009-01-01

Full Text Available Phytoplankton exopolymeric substances (EPS contribute significantly to the dissolved organic car bon (DOC pool in the ocean, playing crucial roles in the surface ocean car bon cycle. Recent studies have demonstrated that ~10% of marine DOC can self-assemble as microgels through electro static Ca bonds providing hotspots of enriched microbial substrate. How ever, the question whether EPS can self-assemble and the formation mechanisms for EPS microgels have not been examined. Here were port that EPS from three representative phytoplankton species, Synechococcus, Emiliania huxleyi, and Skeletonema costatum can spontaneously self assemble in artificial sea water (ASW, forming microscopic gels of ~ 3 - 4 __m in diameter. Different from the marine DOC polymers assembly, these EPS samples can self-assemble in Ca2+-free ASW. Further experiments from fluorescence enhancement and chemical composition analysis confirmed the existence of fair amounts of hydrophobic domains in these EPS samples. These results suggest that hydrophobic interactions play a key role in the assembly of EPS from these three species of marine phytoplankton.

Fuel assemblies for use in nuclear reactors

International Nuclear Information System (INIS)

Mochida, Takaaki.

1987-01-01

Purpose: To increase the plutonium utilization amount and improve the uranium-saving effect in the fuel assemblies of PWR type reactor using mixed uranium-plutonium oxides. Constitution: MOX fuel rods comprising mixed plutonium-uranium oxides are disposed to the outer circumference of a fuel assembly and uranium fuel rods only composed of uranium oxides are disposed to the central portion thereof. In such a fuel assembly, since the uranium fuel rods are present at the periphery of the control rod, the control rod worth is the same as that of the uranium fuel assembly in the prior art. Further, since about 25 % of the entire fuel rods is composed of the MOX fuel rods, the plutonium utilization amount is increased. Further, since the MOX fuel rods at low enrichment degree are present at the outer circumferential portion, mismatching at the boundary to the adjacent MOX fuel assembly is reduced and the problem of local power peaking increase in the MOX fuel assembly is neither present. (Kamimura, M.)

Uranium enrichment management review. Final report

International Nuclear Information System (INIS)

Ellett, J.D.; Rieke, W.B.; Simpson, J.W.; Sullivan, P.E.

1980-01-01

The uranium enrichment enterprise of the US Department of Energy (DOE) provides enriched nuclear fuel for private and government utilities domestically and abroad. The enterprise, in effect, provides a commercial service and represents a signficant business operation within the US government: more than $1 billion in revenues annually and future capital expenditures estimated at several billions of dollars. As a result, in May 1980, the Assistant Secretary for Resource Applications within DOE requested that a group of experienced business executives be assembled to review the operation, financing, and management of the uranium enrichment enterprise as a basis for advising the Secretary of Energy. The review group was specifically asked to focus on the management activities to which sound business practices could be applied. The group developed findings and recommendations in six areas: management of operations and construction; long-range planning; marketing of enrichment services; financial management; research and development; and general management. The chapters of this report present first the management review group's recommendations in the six areas evaluated and then the findings and issues in each area. An appendix provides the group's calendar of meetings. A list of major reference sources used in the course of the study is also included. 12 references

Freeze-Casting Produces a Graphene Oxide Aerogel with a Radial and Centrosymmetric Structure.

Science.gov (United States)

Wang, Chunhui; Chen, Xiong; Wang, Bin; Huang, Ming; Wang, Bo; Jiang, Yi; Ruoff, Rodney S

2018-05-14

We report the assembly of graphene oxide (G-O) building blocks into a vertical and radially aligned structure by a bidirectional freeze-casting approach. The crystallization of water to ice assembles the G-O sheets into a structure, a G-O aerogel whose local structure mimics turbine blades. The centimeter-scale radiating structure in this aerogel has many channels whose width increases with distance from the center. This was achieved by controlling the formation of the ice crystals in the aqueous G-O dispersion that grew radially in the shape of lamellae during freezing. Because the shape and size of ice crystals is influenced by the G-O sheets, different additives (ethanol, cellulose nanofibers, and chitosan) that can form hydrogen bonds with H 2 O were tested and found to affect the interaction between the G-O and formation of ice crystals, producing ice crystals with different shapes. A G-O/chitosan aerogel with a spiral pattern was also obtained. After chemical reduction of G-O, our aerogel exhibited elasticity and absorption capacity superior to that of graphene aerogels with "traditional" pore structures made by conventional freeze-casting. This methodology can be expanded to many other configurations and should widen the use of G-O (and reduced G-O and "graphenic") aerogels.

A new fast neutron collar for safeguards inspection measurements of fresh low enriched uranium fuel assemblies containing burnable poison rods

Energy Technology Data Exchange (ETDEWEB)

Evans, Louise G., E-mail: evanslg@ornl.gov [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Swinhoe, Martyn T.; Menlove, Howard O. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Schwalbach, Peter; Baere, Paul De [European Commission, Euratom Safeguards Office (Luxembourg); Browne, Michael C. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

2013-11-21

Safeguards inspection measurements must be performed in a timely manner in order to detect the diversion of significant quantities of nuclear material. A shorter measurement time can increase the number of items that a nuclear safeguards inspector can reliably measure during a period of access to a nuclear facility. In turn, this improves the reliability of the acquired statistical sample, which is used to inform decisions regarding compliance. Safeguards inspection measurements should also maintain independence from facility operator declarations. Existing neutron collars employ thermal neutron interrogation for safeguards inspection measurements of fresh fuel assemblies. A new fast neutron collar has been developed for safeguards inspection measurements of fresh low-enriched uranium (LEU) fuel assemblies containing gadolinia (Gd{sub 2}O{sub 3}) burnable poison rods. The Euratom Fast Collar (EFC) was designed with high neutron detection efficiency to make a fast (Cd) mode measurement viable whilst meeting the high counting precision and short assay time requirements of the Euratom safeguards inspectorate. A fast mode measurement reduces the instrument sensitivity to burnable poison rod content and therefore reduces the applied poison correction, consequently reducing the dependence on the operator declaration of the poison content within an assembly. The EFC non-destructive assay (NDA) of typical modern European pressurized water reactor (PWR) fresh fuel assembly designs have been simulated using Monte Carlo N-particle extended transport code (MCNPX) simulations. Simulations predict that the EFC can achieve 2% relative statistical uncertainty on the doubles neutron counting rate for a fast mode measurement in an assay time of 600 s (10 min) with the available {sup 241}AmLi (α,n) interrogation source strength of 5.7×10{sup 4} s{sup −1}. Furthermore, the calibration range of the new collar has been extended to verify {sup 235}U content in variable PWR fuel

A new fast neutron collar for safeguards inspection measurements of fresh low enriched uranium fuel assemblies containing burnable poison rods

International Nuclear Information System (INIS)

Evans, Louise G.; Swinhoe, Martyn T.; Menlove, Howard O.; Schwalbach, Peter; Baere, Paul De; Browne, Michael C.

2013-01-01

Safeguards inspection measurements must be performed in a timely manner in order to detect the diversion of significant quantities of nuclear material. A shorter measurement time can increase the number of items that a nuclear safeguards inspector can reliably measure during a period of access to a nuclear facility. In turn, this improves the reliability of the acquired statistical sample, which is used to inform decisions regarding compliance. Safeguards inspection measurements should also maintain independence from facility operator declarations. Existing neutron collars employ thermal neutron interrogation for safeguards inspection measurements of fresh fuel assemblies. A new fast neutron collar has been developed for safeguards inspection measurements of fresh low-enriched uranium (LEU) fuel assemblies containing gadolinia (Gd 2 O 3 ) burnable poison rods. The Euratom Fast Collar (EFC) was designed with high neutron detection efficiency to make a fast (Cd) mode measurement viable whilst meeting the high counting precision and short assay time requirements of the Euratom safeguards inspectorate. A fast mode measurement reduces the instrument sensitivity to burnable poison rod content and therefore reduces the applied poison correction, consequently reducing the dependence on the operator declaration of the poison content within an assembly. The EFC non-destructive assay (NDA) of typical modern European pressurized water reactor (PWR) fresh fuel assembly designs have been simulated using Monte Carlo N-particle extended transport code (MCNPX) simulations. Simulations predict that the EFC can achieve 2% relative statistical uncertainty on the doubles neutron counting rate for a fast mode measurement in an assay time of 600 s (10 min) with the available 241 AmLi (α,n) interrogation source strength of 5.7×10 4 s −1 . Furthermore, the calibration range of the new collar has been extended to verify 235 U content in variable PWR fuel designs in the presence of up to

Gamma counter shutter assembly

International Nuclear Information System (INIS)

Aday, R.W. Jr.; Barber, D.G.

1976-01-01

A shutter assembly for a radioactivity measuring apparatus is described having a sample counting chamber, the assembly having a bulky solid lead cylinder with a sample access port extending therethrough for alignment with the sample chamber. The cylinder is rotated by a Geneva wheel arrangement having a drive wheel with a plurality of equi-angularly disposed pins perpendicular to the surface thereof engaging radially extending open-ended slots in a driven wheel secured to the lead cylinder for concurrent rotation therewith. The drive wheel is rotated at a constant speed with the driven wheel accelerating as a pin traverses the slot from the open end toward the driven wheel center and then decelerating as the pin traverses the reverse direction to provide precise positioning with adjacent pins engaging the open ends of adjacent slots in the stop position of the cylinder. 8 Claims, 3 Drawing Figures

Temperature and boron dependencies of buckling and radial reflector savings for VVER lattices

International Nuclear Information System (INIS)

Alvarez, C.

1990-01-01

The temperature and boron dependencies of buckling and radial reflector savings are analyzed in this paper on the basis of the results from the calculations for the ZR-6M critical assembly. These dependencies are related to he physical behaviour of temperature and boron reactivity coefficients for the cores of VVER-type critical facilities. As a byproduct, the dp/dBg 2 parameter was also investigated and its dependence on water density was determined

Fission blanket benchmark experiment on spherical assembly of uranium and PE with PE reflector

Energy Technology Data Exchange (ETDEWEB)

Zhu, Tonghua; Lu, Xinxin; Wang, Mei; Han, Zijie, E-mail: neutron_integral@aliyun.com; Jiang, Li; Wen, Zhongwei; Liu, Rong

2016-04-15

Highlights: • The fission rate distribution on two depleted uranium assemblies was measured with plate fission chambers. • We do calculations using MCNP code and ENDF/B-V.0 library. • The overestimation of calculations to the measured fission rates was found. • The observed discrepancy are discussed. - Abstract: New concept of fusion-fission hybrid for energy generation has been proposed. To validate the nuclear performance of fission blanket of hybrid, as part of series of validation experiment, two types of fission blanket assemblies were setup in this work and measurements were made of the reaction rate distribution for uranium fission in the spherical assembly of depleted uranium and polyethylene by Plate Fission Chamber (PFC). There are two PFCs in experiment, one is depleted uranium chamber and the other is enriched uranium chamber. The Monte-Carlo transport code MCNP5 and continuous energy cross sections library ENDF/BV.0 were used for the analysis of fission rate distribution in the two types of assemblies. The calculated results were compared with the experimental ones. The overestimation of fission rate for depleted uranium and enriched uranium were found in the inner boundary of the two assemblies. However, the C/E ratio tends to decrease for the distance from the core slightly and the results for enriched uranium are better than that for depleted uranium.

Self-Assembly of Organic Ferroelectrics by Evaporative Dewetting: A Case of β-Glycine.

Science.gov (United States)

Seyedhosseini, Ensieh; Romanyuk, Konstantin; Vasileva, Daria; Vasilev, Semen; Nuraeva, Alla; Zelenovskiy, Pavel; Ivanov, Maxim; Morozovska, Anna N; Shur, Vladimir Ya; Lu, Haidong; Gruverman, Alexei; Kholkin, Andrei L

2017-06-14

Self-assembly of ferroelectric materials attracts significant interest because it offers a promising fabrication route to novel structures useful for microelectronic devices such as nonvolatile memories, integrated sensors/actuators, or energy harvesters. In this work, we demonstrate a novel approach for self-assembly of organic ferroelectrics (as exemplified by ferroelectric β-glycine) using evaporative dewetting, which allows forming quasi-regular arrays of nano- and microislands with preferred orientation of polarization axes. Surprisingly, self-assembled islands are crystallographically oriented in a radial direction from the center of organic "grains" formed during dewetting process. The kinetics of dewetting process follows the t -1/2 law, which is responsible for the observed polygon shape of the grain boundaries and island coverage as a function of radial position. The polarization in ferroelectric islands of β-glycine is parallel to the substrate and switchable under a relatively small dc voltage applied by the conducting tip of piezoresponse force microscope. Significant size effect on polarization is observed and explained within the Landau-Ginzburg-Devonshire phenomenological formalism.

Fuel assembly

International Nuclear Information System (INIS)

Nakajima, Akiyoshi; Bessho, Yasunori; Aoyama, Motoo; Koyama, Jun-ichi; Hirakawa, Hiromasa; Yamashita, Jun-ichi; Hayashi, Tatsuo

1998-01-01

In a fuel assembly of a BWR type reactor in which a water rod of a large diameter is disposed at the central portion, the cross sectional area perpendicular to the axial direction comprises a region a of a fuel rod group facing to a wide gap water region to which a control rod is inserted, a region b of a fuel rod group disposed on the side of the wide gap water region other than the region a, a region d of a fuel rod group facing to a narrow gap water region and a region c of a fuel rod group disposed on the side of the narrow gap water region other than the region d. When comparing an amount of fission products contained in the four regions relative to that in the entire regions and average enrichment degrees of fuel rods for the four regions, the relative amount and the average enrichment degree of the fuel rod group of the region a is minimized, and the relative amount and the average enrichment degree of the fuel rod group in the region b is maximized. Then, reactor shut down margin during cold operation can be improved while flattening the power in the cross section perpendicular to the axial direction. (N.H.)

Watching Nanoscale Self-Assembly Kinetics of Gold Prisms in Liquids

Science.gov (United States)

Kim, Juyeong; Ou, Zihao; Jones, Matthew R.; Chen, Qian

We use liquid-phase transmission electron microscopy to watch self-assembly of gold triangular prisms into polymer-like structures. The in situ dynamics monitoring enabled by liquid-phase transmission electron microscopy, single nanoparticle tracking, and the marked conceptual similarity between molecular reactions and nanoparticle self-assembly combined elucidate the following mechanistic understanding: a step-growth polymerization based assembly statistics, kinetic pathways sampling particle curvature dependent energy minima and their interconversions, and directed assembly into polymorphs (linear or cyclic chains) through in situ modulation of the prism bonding geometry. Our study bridges the constituent kinetics on the molecular and nanoparticle length scales, which enriches the design rules in directed self-assembly of anisotropic nanoparticles.

PWR and BWR spent fuel assembly gamma spectra measurements

Energy Technology Data Exchange (ETDEWEB)

Vaccaro, S. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Tobin, S.J.; Favalli, A. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Hu, J. [Oak Ridge National Laboratory, Oak Ridge (United States); Schwalbach, P. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company (SKB) (Sweden); Trellue, H.; Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States)

2016-10-11

A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of {sup 137}Cs, {sup 154}Eu, and {sup 134}Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

Absorbing rods for nuclear fast neutron reactor absorbing assembly

International Nuclear Information System (INIS)

Aji, M.; Ballagny, A.; Haze, R.

1986-01-01

The invention proposes a neutron absorber rod for neutron absorber assembly of a fast neutron reactor. The assembly comprises a bundle of vertical rods, each one comprising a stack of pellets made of a neutron absorber material contained in a long metallic casing with a certain radial play with regard to this casing; this casing includes traps for splinters from the pellets which may appear during reactor operation, at the level of contact between adjacent pellets. The present invention prevents the casing from rupture involved by the disintegration of the pellets producing pieces of boron carbide of high hardness [fr

Particulate morphology mathematics applied to particle assemblies

CERN Document Server

Gotoh, Keishi

2012-01-01

Encompassing over fifty years of research, Professor Gotoh addresses the correlation function of spatial structures and the statistical geometry of random particle assemblies. In this book morphological study is formed into random particle assemblies to which various mathematics are applied such as correlation function, radial distribution function and statistical geometry. This leads to the general comparison between the thermodynamic state such as gases and liquids and the random particle assemblies. Although structures of molecular configurations change at every moment due to thermal vibration, liquids can be regarded as random packing of particles. Similarly, gaseous states correspond to particle dispersion. If physical and chemical properties are taken away from the subject, the remainder is the structure itself. Hence, the structural study is ubiquitous and of fundamental importance. This book will prove useful to chemical engineers working on powder technology as well as mathematicians interested in le...

The effects of stainless steel radial reflector on core reactivity for small modular reactor

Energy Technology Data Exchange (ETDEWEB)

Kang, Jung Kil, E-mail: jkkang@email.kings.ac.kr; Hah, Chang Joo, E-mail: changhah@kings.ac.kr [KINGS, 658-91, Haemaji-ro, Seosaeng-myeon, Ulju-gun, Ulsan, 689-882 (Korea, Republic of); Cho, Sung Ju, E-mail: sungju@knfc.co.kr; Seong, Ki Bong, E-mail: kbseong@knfc.co.kr [KNFC, Daedeok-daero 989beon-gil, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

2016-01-22

Commercial PWR core is surrounded by a radial reflector, which consists of a baffle and water. Radial reflector is designed to reflect neutron back into the core region to improve the neutron efficiency of the reactor and to protect the reactor vessels from the embrittling effects caused by irradiation during power operation. Reflector also helps to flatten the neutron flux and power distributions in the reactor core. The conceptual nuclear design for boron-free small modular reactor (SMR) under development in Korea requires to have the cycle length of 4∼5 years, rated power of 180 MWth and enrichment less than 5 w/o. The aim of this paper is to analyze the effects of stainless steel radial reflector on the performance of the SMR using UO{sub 2} fuels. Three types of reflectors such as water, water/stainless steel 304 mixture and stainless steel 304 are selected to investigate the effect on core reactivity. Additionally, the thickness of stainless steel and double layer reflector type are also investigated. CASMO-4/SIMULATE-3 code system is used for this analysis. The results of analysis show that single layer stainless steel reflector is the most efficient reflector.

Nuclear fuel assemblies

International Nuclear Information System (INIS)

Natori, Hisahide; Kurihara, Kunitoshi.

1982-01-01

Purpose: To increase the fuel safety by decreasing the gap conductance between fuels and cladding tubes, as well as improve the reactor core controllability by rendering the void coefficient negative. Constitution: Fuel assemblies in a pressure tube comprise a tie-rod, fuel rods in a central region, and fuel rods with burnable poison in the outer circumference region. Here, B 4 C is used as the burnable poison by 1.17 % by weight ratio. The degrees of enrichment for the fissile plutonium as PuO 2 -UO 2 fuel used in the assemblies are 2.7 %, 2.7 % and 1.5 % respectively in the innermost layer, the intermediate layer and the outermost layer. This increases the burn-up degree to improve the plant utilizability, whereby the void coefficient is rendered negative to improve the reactor core controllability. (Horiuchi, T.)

Numerical modelling of the jet nozzle enrichment process

International Nuclear Information System (INIS)

Vercelli, P.

1983-01-01

A numerical model was developed for the simulation of the isotopic enrichment produced by the jet nozzle process. The flow was considered stationary and under ideal gas conditions. The model calculates, for any position of the skimmer piece: (a) values of radial mass concentration profiles for each isotopic species and (b) values of elementary separation effect (Σ sub(A)) and uranium cut (theta). The comparison of the numerical results obtained with the experimental values given in the literature proves the validity of the present work as an initial step in the modelling of the process. (Author) [pt

Template Assembly for Detailed Urban Reconstruction

KAUST Repository

Nan, Liangliang

2015-05-04

We propose a new framework to reconstruct building details by automatically assembling 3D templates on coarse textured building models. In a preprocessing step, we generate an initial coarse model to approximate a point cloud computed using Structure from Motion and Multi View Stereo, and we model a set of 3D templates of facade details. Next, we optimize the initial coarse model to enforce consistency between geometry and appearance (texture images). Then, building details are reconstructed by assembling templates on the textured faces of the coarse model. The 3D templates are automatically chosen and located by our optimization-based template assembly algorithm that balances image matching and structural regularity. In the results, we demonstrate how our framework can enrich the details of coarse models using various data sets.

An improved benchmark model for the Big Ten critical assembly - 021

International Nuclear Information System (INIS)

Mosteller, R.D.

2010-01-01

A new benchmark specification is developed for the BIG TEN uranium critical assembly. The assembly has a fast spectrum, and its core contains approximately 10 wt.% enriched uranium. Detailed specifications for the benchmark are provided, and results from the MCNP5 Monte Carlo code using a variety of nuclear-data libraries are given for this benchmark and two others. (authors)

Critical assembly of uranium enriched to 10% in uranium-235

International Nuclear Information System (INIS)

Hansen, G.E.; Paxton, H.E.

1979-01-01

Big Ten is described in the detail appropriate for a benchmark critical assembly. Characteristics provided are spectral indexes and a detailed neutron flux spectrum, Rossi-α on a reactivity scale established by positive periods, and reactivity coefficients of a variety of isotopes, including the fissionable materials. The observed characteristics are compared with values calculated with ENDF/B-IV cross sections

SNEAK assembly 12B - Construction and experimental program

International Nuclear Information System (INIS)

Helm, F.

1982-10-01

The assembly SNEAK 12A is the second SNEAK assembly in which the neutron physics of disturbed material configurations, as they might occur in a postulated fast reactor accident, is investigated. The core is characterized by a central zone with fuel rods containing PuO 2 UO 2 mixed oxide pellets. This zone is cylindrically surrounded by a buffer zone for spectral adaptation, a driver zone sufficiently reactive to maintain criticality and a radial depleted uranium blanket. The purpose of the experiments is to test calculational methods of neutron physics which are used in the accident analysis of fast breeder reactors. The report describes the construction of the core and its assemblies and their content and outlines the foreseen experimental program. It includes measurements in the clean critical core and investigations of vertical and horizontal fuel relocations and steel blockages

Reducing contrast contamination in radial turbo-spin-echo acquisitions by combining a narrow-band KWIC filter with parallel imaging.

Science.gov (United States)

Neumann, Daniel; Breuer, Felix A; Völker, Michael; Brandt, Tobias; Griswold, Mark A; Jakob, Peter M; Blaimer, Martin

2014-12-01

Cartesian turbo spin-echo (TSE) and radial TSE images are usually reconstructed by assembling data containing different contrast information into a single k-space. This approach results in mixed contrast contributions in the images, which may reduce their diagnostic value. The goal of this work is to improve the image contrast from radial TSE acquisitions by reducing the contribution of signals with undesired contrast information. Radial TSE acquisitions allow the reconstruction of multiple images with different T2 contrasts using the k-space weighted image contrast (KWIC) filter. In this work, the image contrast is improved by reducing the band-width of the KWIC filter. Data for the reconstruction of a single image are selected from within a small temporal range around the desired echo time. The resulting dataset is undersampled and, therefore, an iterative parallel imaging algorithm is applied to remove aliasing artifacts. Radial TSE images of the human brain reconstructed with the proposed method show an improved contrast when compared with Cartesian TSE images or radial TSE images with conventional KWIC reconstructions. The proposed method provides multi-contrast images from radial TSE data with contrasts similar to multi spin-echo images. Contaminations from unwanted contrast weightings are strongly reduced. © 2014 Wiley Periodicals, Inc.

The effect of swelling in Inconel 600 on the performance of FFTF [Fast Flux Test Facility] reflector assemblies

International Nuclear Information System (INIS)

Makenas, B.J.; Trenchard, R.G.; Hecht, S.L.; McCarthy, J.M.; Garner, F.A.

1986-02-01

The Fast Flux Test Facility (FFTF) is designed with non-fueled outer row assemblies, each of which consists of a stack of Inconel 600 blocks penetrated by 316 stainless steel (SS) coolant tubes. These assemblies act as a radial neutron reflector and as a straight but flexible core boundary. During an FFTF refueling outage it was observed that the degree of difficulty in withdrawing an outer row driver fuel assembly was a function of the peak fast fluence of neighboring reflector assemblies. It was subsequently determined through various postirradiation examinations that the reflector assemblies were both bowed and stiff. Measurements of the individual Inconel 600 blocks indicated that the blocks had distorted into a trapezoidal cross section due to differential swelling of Inconel 600 in a steep radial flux gradient. Immersion density results indicate greater irradiation induced volumetric swelling than any previously reported data or correlation for Inconel 600 at equivalent fast fluence. The Inconel 600 swelled approximately the same amount as the SA 316 SS reflector components. Transmission electron microscopy studies on the Inconel blocks and swelling measurements on related materials have been performed and these data have been related to the performance of the reflector materials

Effect of 17 x 17 fuel assembly geometry on interchannel thermal mixing

International Nuclear Information System (INIS)

Motley, F.E.; Wenzell, A.H.; Cadek, F.F.

1975-01-01

A test to determine the value of the thermal diffusion coefficient (TDC) in the 17 x 17 fuel assembly geometry was conducted. The test section was a 5 x 5 rod bundle with a radial power difference of 4.5 to 1. The rod OD and pitch are identical to the 17 x 17 fuel assembly, as is the mixing vane grid design. The value of thermal diffusion coefficient (TDC) was determined by matching the experimental exit enthalpy distribution to that predicted by the THINC computer code. The mean value of TDC for the 17 x 17 fuel assembly geometry is TDC = .059. 6 references

Spatial correction factors for YALINA Booster facility loaded with medium and low enriched fuels

International Nuclear Information System (INIS)

Talamo, A.; Gohar, Y.; Bournos, V.; Fokov, Y.; Kiyavitskaya, H.; Routkovskaya, C.

2012-01-01

The Bell and Glasstone spatial correction factor is used in analyses of subcritical assemblies to correct the experimental reactivity as function of the detector position. Besides the detector position, several other parameters affect the correction factor: the energy weighting function of the detector, the detector size, the energy-angle distribution of source neutrons, and the reactivity of the subcritical assembly. This work focuses on the dependency of the correction factor on the detector material and it investigates the YALINA Booster subcritical assembly loaded with medium (36%) and low (10%) enriched fuels. (authors)

Spatial correction factors for YALINA Booster facility loaded with medium and low enriched fuels

Energy Technology Data Exchange (ETDEWEB)

Talamo, A.; Gohar, Y. [Argonne National Laboratory, 9700 S. Cass Ave, Argonne, IL 60439 (United States); Bournos, V.; Fokov, Y.; Kiyavitskaya, H.; Routkovskaya, C. [Joint Inst. for Power and Nuclear Research-Sosny, 99 Academician A.K.Krasin Str, Minsk 220109 (Belarus)

2012-07-01

The Bell and Glasstone spatial correction factor is used in analyses of subcritical assemblies to correct the experimental reactivity as function of the detector position. Besides the detector position, several other parameters affect the correction factor: the energy weighting function of the detector, the detector size, the energy-angle distribution of source neutrons, and the reactivity of the subcritical assembly. This work focuses on the dependency of the correction factor on the detector material and it investigates the YALINA Booster subcritical assembly loaded with medium (36%) and low (10%) enriched fuels. (authors)

Tungsten--carbide critical assembly

International Nuclear Information System (INIS)

Hansen, G.E.; Paxton, H.C.

1975-06-01

The tungsten--carbide critical assembly mainly consists of three close-fitting spherical shells: a highly enriched uranium shell on the inside, a tungsten--carbide shell surrounding it, and a steel shell on the outside. Ideal critical specifications indicate a rather low computed value of k/sub eff/. Observed and calculated fission-rate distributions for 235 U, 238 U, and 237 Np are compared, and calculated leakage neutrons per fission in various energy groups are given. (U.S.)

Neutronics assessment of thorium-based fuel assembly in SCWR

International Nuclear Information System (INIS)

Liu, Shichang; Cai, Jiejin

2013-01-01

Highlights: • A novel thorium-based fuel assembly for SCWR has been introduced and investigated. • Neutronic properties of three thorium fuels have been studied, compared with UO 2 fuel. • The thorium-based fuel has advantages on fuel utilization and lower MAs generation. -- Abstract: Aiming to take advantage of neutron spectrum of SCWR, a novel thorium-based fuel assembly for SCWR is introduced in this paper. The neutronic characteristics of the introduced fuel assembly with three different thorium fuel types have been investigated using the “dragon” codes. The parameters in different working conditions, such as infinite multiplication factors, radial power peaking factor, temperature coefficient of reactivity and their relation with the operation period have been assessed by comparing with conventional uranium assembly. Moreover, the moderator-to-fuel ratio (MFR) was changed in order to investigate its influence on the neutronic characteristics of fuel assembly. Results show that the thorium-based fuel has advantages on both efficient fuel utilization and lower minor actinide generation, with some similar neutronic properties to the uranium fuel

Verification of maximum radial power peaking factor due to insertion of FPM-LEU target in the core of RSG-GAS reactor

Energy Technology Data Exchange (ETDEWEB)

Setyawan, Daddy, E-mail: d.setyawan@bapeten.go.id [Center for Assessment of Regulatory System and Technology for Nuclear Installations and Materials, Indonesian Nuclear Energy Regulatory Agency (BAPETEN), Jl. Gajah Mada No. 8 Jakarta 10120 (Indonesia); Rohman, Budi [Licensing Directorate for Nuclear Installations and Materials, Indonesian Nuclear Energy Regulatory Agency (BAPETEN), Jl. Gajah Mada No. 8 Jakarta 10120 (Indonesia)

2014-09-30

Verification of Maximum Radial Power Peaking Factor due to insertion of FPM-LEU target in the core of RSG-GAS Reactor. Radial Power Peaking Factor in RSG-GAS Reactor is a very important parameter for the safety of RSG-GAS reactor during operation. Data of radial power peaking factor due to the insertion of Fission Product Molybdenum with Low Enriched Uranium (FPM-LEU) was reported by PRSG to BAPETEN through the Safety Analysis Report RSG-GAS for FPM-LEU target irradiation. In order to support the evaluation of the Safety Analysis Report incorporated in the submission, the assessment unit of BAPETEN is carrying out independent assessment in order to verify safety related parameters in the SAR including neutronic aspect. The work includes verification to the maximum radial power peaking factor change due to the insertion of FPM-LEU target in RSG-GAS Reactor by computational method using MCNP5and ORIGEN2. From the results of calculations, the new maximum value of the radial power peaking factor due to the insertion of FPM-LEU target is 1.27. The results of calculations in this study showed a smaller value than 1.4 the limit allowed in the SAR.

10 CFR Appendix B to Part 110 - Illustrative List of Gas Centrifuge Enrichment Plant Components Under NRC's Export Licensing...

Science.gov (United States)

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Illustrative List of Gas Centrifuge Enrichment Plant... 110—Illustrative List of Gas Centrifuge Enrichment Plant Components Under NRC's Export Licensing Authority 1. Assemblies and components especially designed or prepared for use in gas centrifuges. Note: The...

Low-enriched research reactor fuel: Post-Irradiation Examinations at SCK-CEN

International Nuclear Information System (INIS)

Van den Berghe, S.; Leenaers, A.

2007-01-01

Generally, research and test reactors are fuelled with fuel plates instead of pins. In most cases in the past, these plates consisted of high enriched (higher than 95 percent 235 U) UAl 3 powder mixed with a pure Al matrix (called the meat) in between two aluminium alloy plates (the cladding). These plates are then assembled in fuel elements of different designs to fit the needs of the various reactors. Since the 1970's, efforts have been going on to replace the high-enriched, low-density UAl 3 fuel with high-density, low enriched ( 235 U) replacements. This search is driven by the attempt to reduce the civil use of high-enriched materials because of proliferation risks and terrorist threats. American initiatives, such as the Global Threat Reduction Initiative and the Reduced Enrichment for Research and Test Reactors program have triggered the development of reliable low-enriched fuel types for these reactors, which can replace the high enriched ones without loss of performance. Most success has been obtained with U 3 Si 2 fuel, which is currently used in many research reactors in the world. However, efforts to search for a better replacement have continued and are currently directed towards the U-Mo alloy fuel (7-10 weight percent Mo)

Antiproton compression and radial measurements

CERN Document Server

Andresen, G B; Bowe, P D; Bray, C C; Butler, E; Cesar, C L; Chapman, S; Charlton, M; Fajans, J; Fujiwara, M C; Funakoshi, R; Gill, D R; Hangst, J S; Hardy, W N; Hayano, R S; Hayden, M E; Humphries, A J; Hydomako, R; Jenkins, M J; Jorgensen, L V; Kurchaninov, L; Lambo, R; Madsen, N; Nolan, P; Olchanski, K; Olin, A; Page R D; Povilus, A; Pusa, P; Robicheaux, F; Sarid, E; Seif El Nasr, S; Silveira, D M; Storey, J W; Thompson, R I; Van der Werf, D P; Wurtele, J S; Yamazaki, Y

2008-01-01

Control of the radial profile of trapped antiproton clouds is critical to trapping antihydrogen. We report detailed measurements of the radial manipulation of antiproton clouds, including areal density compressions by factors as large as ten, achieved by manipulating spatially overlapped electron plasmas. We show detailed measurements of the near-axis antiproton radial profile, and its relation to that of the electron plasma. We also measure the outer radial profile by ejecting antiprotons to the trap wall using an octupole magnet.

Critical and shielding parametric studies with the Monte Carlo code TRIPOLI to identify the key points to take into account during the transportation of blanket assemblies with high ratio of americium

International Nuclear Information System (INIS)

Gosmain, Cecile-Aline

2011-01-01

In the framework of French research program on Generation IV sodium cooled fast reactor, one possible option consists in burning minor actinides in this kind of Advanced Sodium Technological Reactor. Two types of transmutation mode are studied in the world : the homogeneous mode of transmutation where actinides are scattered with very low enrichment ratio in fissile assemblies and the heterogeneous mode where fissile core is surrounded by blanket assemblies filled with minor actinides with ratio of incorporated actinides up to 20%. Depending on which element is considered to be burnt and on its content, these minor actinides contents imply constraints on assemblies' transportation between Nuclear Power Plants and fuel cycle facilities. In this study, we present some academic studies in order to identify some key constraints linked to the residual power and neutron/gamma load of such kind of blanket assemblies. To simplify the approach, we considered a modeling of a 'model cask' dedicated to the transportation of a unique irradiated blanket assembly loaded with 20% of Americium and basically inspired from an existent cask designed initially for the damaged fissile Superphenix assembly transport. Thermal calculations performed with EDF-SYRTHES code have shown that due to thermal limitations on cladding temperature, the decay time to be considered before transportation is 20 years. This study is based on explicit 3D representations of the cask and the contained blanket assembly with the Monte Carlo code TRIPOLI/JEFF3.1.1 library and concludes that after such a decay time, the transportation of a unique Americium radial blanket is feasible only if the design of our model cask is modified in order to comply with the dose limitation criterion. (author)

Calculation of the fissile mass of a graphite moderated critical assembly using 93% enriched uranium

International Nuclear Information System (INIS)

Correa, F.; Marzo, M.A.S.; Collussi, I.; Ferreira, A.C.A.

1976-01-01

The critical mass of uranium has been calculated for a graphite moderated set fueled with 93% enriched uranium to be mounted on the Instituto de Energia Atomica split table Zero Power Reactor. The core composition was optimized to permit the maximum number of configurations to be studied. Analysis of three core compositions shows that 8 Kg of uranium enriched to 93% - U-235 (by weight) and 100 Kg of thorium would be sufficient for criticality experiments [pt

Georges Besse 2. A new era for enrichment

International Nuclear Information System (INIS)

2009-01-01

Since 1978, AREVA group subsidiary EURODIF's Georges Besse plant has been using gaseous diffusion to enrich uranium and meet the requirements of electricity utilities. Georges Besse II plant, which will use far less electricity, will replace George Besse not so long. The Georges Besse II plant is located on the Tricastin nuclear site in southern France. AREVA ensure delivery of uranium enrichment services in accordance with customer expectations. AREVA obtained the right to use URENCO's centrifugation technology on July 3, 2006. This is the process to be used at Georges Besse II. The new uranium enrichment plant will comprise two enrichment units, with a total production capacity of 7.5 million separative work units, which can be extended to 11 million if needed. Each enrichment unit will include: a Centrifuge Assembly Building (CAB), a Centrifuge Utility Building (CUB) with offices and control room, annexes for purification, supply and extraction of uranium hexafluoride (UF 6 ), modules containing the halls that house the centrifuge cascades. The modular design of the Georges Besse II plant will allow production in the first cascade where as the others cascades are build. The first cascade should be operational in the first half of 2009. At an overall cost of 3 billion euros, this project is one of the largest investments of the decade in France On November 24, 2003, AREVA and URENCO signed an agreement under which AREVA would buy 50% of the shares in the Enrichment Technology Company (ETC), which designs and manufactures centrifuges

Stability of radial swirl flows

International Nuclear Information System (INIS)

Dou, H S; Khoo, B C

2012-01-01

The energy gradient theory is used to examine the stability of radial swirl flows. It is found that the flow of free vortex is always stable, while the introduction of a radial flow will induce the flow to be unstable. It is also shown that the pure radial flow is stable. Thus, there is a flow angle between the pure circumferential flow and the pure radial flow at which the flow is most unstable. It is demonstrated that the magnitude of this flow angle is related to the Re number based on the radial flow rate, and it is near the pure circumferential flow. The result obtained in this study is useful for the design of vaneless diffusers of centrifugal compressors and pumps as well as other industrial devices.

Measuring magnetic correlations in nanoparticle assemblies

DEFF Research Database (Denmark)

Beleggia, Marco; Frandsen, Cathrine

2014-01-01

We illustrate how to extract correlations between magnetic moments in assemblies of nanoparticles from, e.g., electron holography data providing the combined knowledge of particle size distribution, inter-particle distances, and magnitude and orientation of each magnetic moment within...... a nanoparticle superstructure, We show, based on simulated data, how to build a radial/angular pair distribution function f(r,θ) encoding the spatial and angular difference between every pair of magnetic moments. A scatter-plot of f(r,θ) reveals the degree of structural and magnetic order present, and hence...

Criticality experiments with low enriched UO2 fuel rods in water containing dissolved gadolinium

International Nuclear Information System (INIS)

Bierman, S.R.; Murphy, E.S.; Clayton, E.D.; Keay, R.T.

1984-02-01

The results obtained in a criticality experiments program performed for British Nuclear Fuels, Ltd. (BNFL) under contract with the United States Department of Energy (USDOE) are presented in this report along with a complete description of the experiments. The experiments involved low enriched UO 2 and PuO 2 -UO 2 fuel rods in water containing dissolved gadolinium, and are in direct support of BNFL plans to use soluble compounds of the neutron poison gadolinium as a primary criticality safeguard in the reprocessing of low enriched nuclear fuels. The experiments were designed primarily to provide data for validating a calculation method being developed for BNFL design and safety assessments, and to obtain data for the use of gadolinium as a neutron poison in nuclear chemical plant operations - particularly fuel dissolution. The experiments program covers a wide range of neutron moderation (near optimum to very under-moderated) and a wide range of gadolinium concentration (zero to about 2.5 g Gd/l). The measurements provide critical and subcritical k/sub eff/ data (1 greater than or equal to k/sub eff/ greater than or equal to 0.87) on fuel-water assemblies of UO 2 rods at two enrichments (2.35 wt % and 4.31 wt % 235 U) and on mixed fuel-water assemblies of UO 2 and PuO 2 -UO 2 rods containing 4.31 wt % 235 U and 2 wt % PuO 2 in natural UO 2 respectively. Critical size of the lattices was determined with water containing no gadolinium and with water containing dissolved gadolinium nitrate. Pulsed neutron source measurements were performed to determine subcritical k/sub eff/ values as additional amounts of gadolinium were successively dissolved in the water of each critical assembly. Fission rate measurements in 235 U using solid state track recorders were made in each of the three unpoisoned critical assemblies, and in the near-optimum moderated and the close-packed poisoned assemblies of this fuel

[Comparison of chemical quality characteristics between radial striations and non-radial striations in tuberous root of Rehmannia glutinosa].

Science.gov (United States)

Xie, Cai-Xia; Zhang, Miao; Li, Ya-Jing; Geng, Xiao-Tong; Wang, Feng-Qing; Zhang, Zhong-Yi

2017-11-01

An HPLC method was established to determine the contents of catalpol, acteoside, rehmaionoside A, rehmaionoside D, leonuride in three part of Rehmanni glutinosa in Beijing No.1 variety R. glutinosa during the growth period, This method, in combination with its HPLC fingerprint was used to evaluate its overall quality characteristics.The results showed that：① the content of main components of R. glutinosa varied in different growth stages ;② there was a great difference of the content of main components between theradial striations and the non-radial striations; ③ the two sections almost have the same content distribution of catalpol, acteoside and rehmaionoside D; ④the content of rehmaionoside A in non-radial striations was higher than that in radial striations,while the content of leonuride in radial striations was higher than that in non-radial striations.; ⑤the HPLC fingerprint of radial striations, non-radial striations and whole root tuber were basically identical, except for the big difference in the content of chemical components. The result of clustering displayed that the radial striations, non-radial striations, and whole root were divided into two groups. In conclusion, there was a significant difference in the quality characteristics of radial striations and non-radial striations of R. glutinosa. This research provides a reference for quality evaluation and geoherbalism of R. glutinosa. Copyright© by the Chinese Pharmaceutical Association.

Evaluation of the radial design of fuel cells in an operation cycle of a BWR reactor; Evaluacion del diseno radial de celdas de combustible en un ciclo de operacion de un reactor BWR

Energy Technology Data Exchange (ETDEWEB)

Gonzalez C, J.; Martin del Campo M, C. [Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Facultad de Ingenieria, UNAM, Paseo Cuauhnahuac 8532, Jiutepec, Morelos (Mexico)]. e-mail: jgco@ver.megared.net.mx

2003-07-01

This work is continuation of one previous in the one that the application of the optimization technique called Tabu search to the radial design of fuel cells of boiling water reactors (BWR, Boiling Water Reactor) is presented. The objective function used in the optimization process only include neutron parameters (k-infinite and peak of radial power) considering the cell at infinite media. It was obtained to reduce the cell average enrichment completing the characteristics of reactivity of an original cell. The objective of the present work is to validate the objective function that was used for the radial design of the fuel cell (test cell), analyzing the operation of a one cycle of the reactor in which fuels have been fresh recharged that contain an axial area with the nuclear database of the cell designed instead of the original cell. For it is simulated it with Cm-Presto the cycle 10 of the reactor operation of the Unit 1 of the Nuclear Power station of Laguna Verde (U1-CNLV). For the cycle evaluation its were applied so much the simulation with the Haling strategy, as the simulation of the one cycle with control rod patterns and they were evaluated the energy generation and several power limits and reactivity that are used as design parameters in fuel reloads of BWR reactors. The results at level of an operation cycle of the reactor, show that the objective function used in the optimization and radial design of the cell is adequate and that it can induce to one good use of the fuel. (Author)

Experimental feasibility study of radial injection cooling of three-pad radial air foil bearings

Science.gov (United States)

Shrestha, Suman K.

Air foil bearings use ambient air as a lubricant allowing environment-friendly operation. When they are designed, installed, and operated properly, air foil bearings are very cost effective and reliable solution to oil-free turbomachinery. Because air is used as a lubricant, there are no mechanical contacts between the rotor and bearings and when the rotor is lifted off the bearing, near frictionless quiet operation is possible. However, due to the high speed operation, thermal management is one of the very important design factors to consider. Most widely accepted practice of the cooling method is axial cooling, which uses cooling air passing through heat exchange channels formed underneath the bearing pad. Advantage is no hardware modification to implement the axial cooling because elastic foundation structure of foil bearing serves as a heat exchange channels. Disadvantage is axial temperature gradient on the journal shaft and bearing. This work presents the experimental feasibility study of alternative cooling method using radial injection of cooling air directly on the rotor shaft. The injection speeds, number of nozzles, location of nozzles, total air flow rate are important factors determining the effectiveness of the radial injection cooling method. Effectiveness of the radial injection cooling was compared with traditional axial cooling method. A previously constructed test rig was modified to accommodate a new motor with higher torque and radial injection cooling. The radial injection cooling utilizes the direct air injection to the inlet region of air film from three locations at 120° from one another with each location having three axially separated holes. In axial cooling, a certain axial pressure gradient is applied across the bearing to induce axial cooling air through bump foil channels. For the comparison of the two methods, the same amount of cooling air flow rate was used for both axial cooling and radial injection. Cooling air flow rate was

Power peak in vicinity of WWER-440 control assembly

International Nuclear Information System (INIS)

Mikus, J.

2002-01-01

This paper presents information concerning the WWER-440 local power peaking problem induced by a control assembly and corresponding investigation possibilities on the light-water zero-power reactor LR-O at the Nuclear Research Institute Rez plc. Brief description of the disposable CA model, experimental arrangement and conditions on the LR-O reactor, preparation of the relevant measurements in the WWER-440 type cores with CA model, as well as some preliminary results of the fission density distribution obtained in a core without boron and with fuel assemblies having profiled enrichment are mentioned too (Author)

Physical mechanism determining the radial electric field and its radial structure in a toroidal plasma

International Nuclear Information System (INIS)

Ida, Katsumi; Miura, Yukitoshi; Itoh, Sanae

1994-10-01

Radial structures of plasma rotation and radial electric field are experimentally studied in tokamak, heliotron/torsatron and stellarator devices. The perpendicular and parallel viscosities are measured. The parallel viscosity, which is dominant in determining the toroidal velocity in heliotron/torsatron and stellarator devices, is found to be neoclassical. On the other hand, the perpendicular viscosity, which is dominant in dictating the toroidal rotation in tokamaks, is anomalous. Even without external momentum input, both a plasma rotation and a radial electric field exist in tokamaks and heliotrons/torsatrons. The observed profiles of the radial electric field do not agree with the theoretical prediction based on neoclassical transport. This is mainly due to the existence of anomalous perpendicular viscosity. The shear of the radial electric field improves particle and heat transport both in bulk and edge plasma regimes of tokamaks. (author) 95 refs

WWER-440 fuel cycles possibilities using improved fuel assemblies design

International Nuclear Information System (INIS)

Mikolas, P.; Svarny, J.

2008-01-01

Practically five years cycle has been achieved in the last years at NPP Dukovany. There are two principal means how it could be achieved. First, it is necessary to use fuel assemblies with higher fuel enrichment and second, to use fuel loading with very low leakage. Both these conditions are fulfilled at NPP Dukovany at this time. It is known, that the fuel cycle economy can be improved by increasing the fuel residence time in the core up to six years. There are at least two ways how this goal could be achieved. The simplest way is to increase enrichment in fuel. There exists a limit, which is 5.0 w % of 235 U. Taking into account some uncertainty, the calculation maximum is 4.95 w % of 235 U. The second way is to change fuel assembly design. There are several possibilities, which seem to be suitable from the neutron - physical point of view. The first one is higher mass content of uranium in a fuel assembly. The next possibility is to enlarge pin pitch. The last possibility is to 'omit' FA shroud. This is practically unrealistic; anyway, some other structural parts must be introduced. The basic neutron physical characteristics of these cycles for up-rated power are presented showing that the possibilities of fuel assemblies with this improved design in enlargement of fuel cycles are very promising. In the end, on the basis of neutron physical characteristics and necessary economical input parameters, a preliminary evaluation of economic contribution of proposals of advanced fuel assemblies on fuel cycle economy is presented (Authors)

Search Greedy for radial fuel optimization; Busqueda Greddy para optimizacion radial de combustible

Energy Technology Data Exchange (ETDEWEB)

Ortiz, J. J.; Castillo, J. A. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico); Pelta, D. A. [Universidad de Granada, ETS Ingenieria Informatica y Telecomunicaciones, C/Daniel Saucedo Aranda s/n, 18071 Granada (Spain)]. e-mail: jjortiz@nuclear.inin.mx

2008-07-01

In this work a search algorithm Greedy is presented for the optimization of fuel cells in reactors BWR. As first phase a study was made of sensibility of the Factor of Pick of Local Power (FPPL) of the cell, in function of the exchange of the content of two fuel rods. His way it could settle down that then the rods to exchange do not contain gadolinium, small changes take place in the value of the FPPL of the cell. This knowledge was applied later in the search Greedy to optimize fuel cell. Exchanges of rods with gadolinium takes as a mechanism of global search and exchanges of rods without gadolinium takes as a method of local search. It worked with a cell of 10x10 rods and 2 circular water channels in center of the same one. From an inventory of enrichments of uranium and concentrations of given gadolinium and one distribution of well-known enrichments; the techniques finds good solutions that the FPPL minimizes, maintaining the factor of multiplication of neutrons in a range appropriate of values. In the low part of the assembly of a lot of recharge of a cycle of 18 months the cells were placed. The values of FPPL of the opposing cells are similar or smaller to those of the original cell and with behaviors in the nucleus also comparable to those obtained with the original cell. The evaluation of the cells was made with the code of transport CASMO-IV and the evaluation of the nucleus was made by means of the one simulator of the nucleus SIMULATE-3. (Author)

Radionuclide inventories : ORIGEN2.2 isotopic depletion calculation for high burnup low-enriched uranium and weapons-grade mixed-oxide pressurized-water reactor fuel assemblies.

Energy Technology Data Exchange (ETDEWEB)

Gauntt, Randall O.; Ross, Kyle W. (Los Alamos National Laboratory, Los Alamos, NM); Smith, James Dean; Longmire, Pamela

2010-04-01

The Oak Ridge National Laboratory computer code, ORIGEN2.2 (CCC-371, 2002), was used to obtain the elemental composition of irradiated low-enriched uranium (LEU)/mixed-oxide (MOX) pressurized-water reactor fuel assemblies. Described in this report are the input parameters for the ORIGEN2.2 calculations. The rationale for performing the ORIGEN2.2 calculation was to generate inventories to be used to populate MELCOR radionuclide classes. Therefore the ORIGEN2.2 output was subsequently manipulated. The procedures performed in this data reduction process are also described herein. A listing of the ORIGEN2.2 input deck for two-cycle MOX is provided in the appendix. The final output from this data reduction process was three tables containing the radionuclide inventories for LEU/MOX in elemental form. Masses, thermal powers, and activities were reported for each category.

Fuel cycles of WWER-1000 based on assemblies with increased fuel mass

International Nuclear Information System (INIS)

Kosourov, E.; Pavlovichev, A.; Shcherenko, A.

2011-01-01

Modern WWER-1000 fuel cycles are based on FAs with the fuel column height of 3680 mm, diameters of the fuel pellet and its central hole of 7.6 and 1.2 mm respectively. The highest possible fuel enrichment has reached its license limit that is 4.95 %. Research in the field of modernization, safety justification and licensing of equipment for fuel manufacture, storage and transportation are required for further fuel enrichment increase (above 5 %). So in the nearest future an improvement of technical and economic characteristics of fuel cycles is possible if assembly fuel mass is increased. The available technology of the cladding thinning makes it possible. If the fuel rod outer diameter is constant and the clad inner diameter is increased to 7.93 mm, the diameter of the fuel pellet can be increased to 7.8 mm. So the suppression of the pellet central hole allows increasing assembly fuel weight by about 8 %. In this paper we analyze how technical and economic characteristics of WWER-1000 fuel cycle change when an advanced FA is applied instead of standard one. Comparison is made between FAs with equal time interval between refueling. This method of comparison makes it possible to eliminate the parameters that constitute the operation component of electricity generation cost, taking into account only the following technical and economic characteristics: 1)cycle length; 2) average burnup of spent FAs; 3) specific natural uranium consumption; 4)specific quantity of separative work units; 5) specific enriched uranium consumption; 6) specific assembly consumption. Collected data allow estimating the efficiency of assembly fuel weight increase and verifying fuel cycle characteristics that may be obtained in the advanced FAs. (authors)

Calibration of spent fuel measurement assembly

International Nuclear Information System (INIS)

Koleska, Michal; Viererbl, Ladislav; Marek, Milan

2014-01-01

The LVR-15 research reactor (Czech Republic) had been converted from the highly enriched IRT-2M to the low enriched IRT-4M fuel. For the possibility of the independent pre-transport evaluation of IRT-2M burnup, a spectrometric system was developed. This spectrometric system consists of the fuel holder, the collimator and the portable Canberra Big MAC HPGe (High Purity Germanium) detector. In order to have well reproducible and reliable experimental data for modeling of the measurement system, calibration with the 110m Ag isotope with known activity was performed. This isotope was chosen for having energies similar to isotopes measured in fuel assemblies. The 110m Ag isotope was prepared by irradiating of the silver foil in LVR-15 research reactor; its activity was evaluated in the LVR-15's spectrometric laboratory. From the measured data, an efficiency curve of the spectrometric system has been determined. The experimental data were compared to the calculation results with the MCNPX model of the spectrometric system. - Highlights: • Calibration of research reactor spent fuel measurement assembly. • On-site prepared 110m Ag isotope used for the measurement. • Calculated self-shielding factor for the IRT-2M fuel. • Applicable to other research reactor fuel geometries

Development of ORIGEN libraries for mixed oxide (MOX) fuel assembly designs

International Nuclear Information System (INIS)

Mertyurek, Ugur; Gauld, Ian C.

2016-01-01

Highlights: • ORIGEN MOX library generation process is described. • SCALE burnup calculations are validated against measured MOX fuel samples from the MALIBU program. • ORIGEN MOX libraries are verified using the OECD Phase IV-B benchmark. • There is good agreement for calculated-to-measured isotopic distributions. - Abstract: ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup. The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. The nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.

An estimate of the reactivity of assemblies of NRX fuel elements in light water

International Nuclear Information System (INIS)

Jarvis, R.G.

1960-03-01

This report contains calculations on the criticality of assemblies of NRX fuel elements in light water. The elements are dealt with in three sections, 'X rods' of natural uranium, enriched elements of U 235 /A1 alloy and enriched elements of Pu/Al alloy. Values of k ∞ and B 2 are provided for two fuel concentrations for each of the two enriched types and for a range of irradiations of the X rods. The calculations for the X rods provide maximum and minimum values of k ∞ . The maximum values for some lattices are a few per cent above unity. Unfortunately, the present experimental evidence does not prove that it is impossible to achieve values of k ∞ greater than unity in lattices of natural uranium in light water. Hence for safety predictions maximum values have been used. The resulting restrictions are not very severe. It is possible to make critical assemblies of the enriched elements, Part (5) contains a set of recommended minimum spacings such that elements of all kinds may safely be mixed in a stack together. There are also predictions of the minimum critical numbers of complete elements or elements cut into slugs. (author)

Radial retinotomy in the macula.

Science.gov (United States)

Bovino, J A; Marcus, D F

1984-01-01

Radial retinotomy is an operative procedure usually performed in the peripheral or equatorial retina. To facilitate retinal attachment, the authors used intraocular scissors to perform radial retinotomy in the macula of two patients during vitrectomy surgery. In the first patient, a retinal detachment complicated by periretinal proliferation and macula hole formation was successfully reoperated with the aid of three radial cuts in the retina at the edges of the macular hole. In the second patient, an intraoperative retinal tear in the macula during diabetic vitrectomy was also successfully repaired with the aid of radial retinotomy. In both patients, retinotomy in the macula was required because epiretinal membranes, which could not be easily delaminated, were hindering retinal reattachment.

Radial head dislocation during proximal radial shaft osteotomy.

Science.gov (United States)

Hazel, Antony; Bindra, Randy R

2014-03-01

The following case report describes a 48-year-old female patient with a longstanding both-bone forearm malunion, who underwent osteotomies of both the radius and ulna to improve symptoms of pain and lack of rotation at the wrist. The osteotomies were templated preoperatively. During surgery, after performing the planned radial shaft osteotomy, the authors recognized that the radial head was subluxated. The osteotomy was then revised from an opening wedge to a closing wedge with improvement of alignment and rotation. The case report discusses the details of the operation, as well as ways in which to avoid similar shortcomings in the future. Copyright © 2014 American Society for Surgery of the Hand. Published by Elsevier Inc. All rights reserved.

U.S. Commercial Spent Nuclear Fuel Assembly Characteristics - 1968-2013

Energy Technology Data Exchange (ETDEWEB)

Hu, Jianwei [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Peterson, Joshua L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bowman, Stephen M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-09-01

Activities related to management of spent nuclear fuel (SNF) are increasing in the US and many other countries. Over 240,000 SNF assemblies have been discharged from US commercial reactors since the late 1960s. The enrichment and burnup of SNF have changed significantly over the past 40 years, and fuel assembly designs have also evolved. Understanding the general characteristics of SNF helps regulators and other stakeholders form overall strategies towards the final disposal of US SNF. This report documents a survey of all US commercial SNF assemblies in the GC-859 database and provides reference SNF source terms (e.g., nuclide inventories, decay heat, and neutron/photon emission) at various cooling times up to 200 years after fuel discharge. This study reviews the distribution and evolution of fuel parameters of all SNF assemblies discharged over the past 40 years. Assemblies were categorized into three groups based on discharge year, and the median burnups and enrichments of each group were used to establish representative cases. An extended burnup case was created for boiling water reactor (BWR) fuels, and another was created for the pressurized water reactor (PWR) fuels. Two additional cases were developed to represent the eight mixed oxide (MOX) fuel assemblies in the database. Burnup calculations were performed for each representative case. Realistic parameters for fuel design and operations were used to model the SNF and to provide reference fuel characteristics representative of the current inventory. Burnup calculations were performed using the ORIGEN code, which is part of the SCALE nuclear modeling and simulation code system. Results include total activity, decay heat, photon emission, neutron flux, gamma heat, and plutonium content, as well as concentrations for 115 significant nuclides. These quantities are important in the design, regulation, and operations of SNF storage, transportation, and disposal systems.

Reconstitutable control rod spider assembly

International Nuclear Information System (INIS)

Shallenberger, J.M.; Ferian, S.J.

1990-01-01

A reconstitutable control rod/spider assembly includes a hollow connecting finger of the spider having a pair of opposing flat segments formed on the interior thereof and engaging a pair of opposing flat sectors formed on the exterior of a stem extending form the upper end of control rod. The stem also has an externally-threaded portion engaging a nut and a pilot aligning portion for the nut. The nut has a radially flexible and expandable thread-defining element captured in its bore. The segments and sectors allow the rod to be removed and reattached after turning through 180 0 to allow more even wear on the rod. (author)

Thermal-hydraulic and neutron-physical characteristics of a new SCWR fuel assembly

International Nuclear Information System (INIS)

Liu, X.J.; Cheng, X.

2009-01-01

A new fuel assembly design for a thermal supercritical water cooled reactor (SCWR) core is proposed. Compared to the existing fuel assemblies, the present fuel assembly has two-rows of fuel rods between the moderator channels, to achieve a more uniform moderation for all fuel rod cells, and subsequently, a more uniform radial power distribution. In addition, a neutron-kinetics/thermal-hydraulics coupling method is developed, to analyze the neutron-physical and thermal-hydraulic behavior of the fuel assembly designs. This coupling method is based on the sub-channel analysis code COBRA-IV for thermal-hydraulics and the neutron-kinetics code SKETCH-N for neutron-physics. Both the COBRA-IV code and the SKETCH-N code are accordingly modified. An interface is established for the data transfer between these two codes. This coupling method is applied to both the one-row fuel assemblies (previous design) and the two-row fuel assemblies (present design). The performance of the two types of fuel assemblies is compared. The results show clearly that the two-row fuel assembly has more favorable neutron-physical and thermal-hydraulic characteristics than the one-row fuel assembly. The effect of various parameters on the fuel assembly performance is discussed. The coupling method is proven to be well suitable for further applications to SCWR fuel assembly design analysis

Radial cracks and fracture mechanism of radially oriented ring 2:17 type SmCo magnets

International Nuclear Information System (INIS)

Tian Jianjun; Pan Dean; Zhou Hao; Yin Fuzheng; Tao Siwu; Zhang Shengen; Qu Xuanhui

2009-01-01

Radially oriented ring 2:17 type SmCo magnets have different microstructure in the radial direction (easy magnetization) and axial direction (hard magnetization). The structure of the cross-section in radial direction is close-packed atomic plane, which shows cellular microstructure. The microstructure of the cross-section in axial direction consists of a mixture of rhombic microstructure and parallel lamella phases. So the magnets have obvious anisotropy of thermal expansion in different directions. The difference of the thermal expansion coefficients reaches the maximum value at 830-860 deg. C, which leads to radial cracks during quenching. The magnets have high brittlement because there are fewer slip systems in crystal structure. The fracture is brittle cleavage fracture.

Design and research of seal structure for thermocouple column assembly

International Nuclear Information System (INIS)

Rao Qiqi; Li Na; Zhao Wei; Ma Zhigang

2015-01-01

The new seal structure was designed to satisfy the function of thermocouple column assembly and the reactor structure. This seal structure uses the packing graphite ring and adopts the self-sealing principle. Cone angle is brought to the seal face of seal structure which is conveniently to assembly and disassembly. After the sealing principle analysis and stress calculation of graphite ring which adopt the cone angle, the cone angle increases the radial force of seal structure and improves the seal effect. The stress analysis result shows the seal structure strength satisfies the regulation requirement. The cold and hot function test results shows the sealing effect is good, and the design requirement is satisfied. (authors)

Yeast prions assembly and propagation: contributions of the prion and non-prion moieties and the nature of assemblies.

Science.gov (United States)

Kabani, Mehdi; Melki, Ronald

2011-01-01

Yeast prions are self-perpetuating protein aggregates that are at the origin of heritable and transmissible non-Mendelian phenotypic traits. Among these, [PSI+], [URE3] and [PIN+] are the most well documented prions and arise from the assembly of Sup35p, Ure2p and Rnq1p, respectively, into insoluble fibrillar assemblies. Fibril assembly depends on the presence of N- or C-terminal prion domains (PrDs) which are not homologous in sequence but share unusual amino-acid compositions, such as enrichment in polar residues (glutamines and asparagines) or the presence of oligopeptide repeats. Purified PrDs form amyloid fibrils that can convert prion-free cells to the prion state upon transformation. Nonetheless, isolated PrDs and full-length prion proteins have different aggregation, structural and infectious properties. In addition, mutations in the "non-prion" domains (non-PrDs) of Sup35p, Ure2p and Rnq1p were shown to affect their prion properties in vitro and in vivo. Despite these evidences, the implication of the functional non-PrDs in fibril assembly and prion propagation has been mostly overlooked. In this review, we discuss the contribution of non-PrDs to prion assemblies, and the structure-function relationship in prion infectivity in the light of recent findings on Sup35p and Ure2p assembly into infectious fibrils from our laboratory and others.

VizieR Online Data Catalog: Li enrichment histories of the thick/thin disc (Fu+, 2018)

Science.gov (United States)

Fu, X.; Romano, D.; Bragaglia, A.; Mucciarelli, A.; Lind, K.; Delgado Mena, E.; Sousa, S. G.; Randich, S.; Bressan, A.; Sbordone, L.; Martell, S.; Korn, A. J.; Abia, C.; Smiljanic, R.; Jofre, P.; Pancino, E.; Tautvaisiene, G.; Tang, B.; Lanzafame, A. C.; Magrini, L.; Carraro, G.; Bensby, T.; Damiani, F.; Alfaro, E. J.; Flaccomio, E.; Morbidelli, L.; Zaggia, S.; Lardo, C.; Monaco, L.; Frasca, A.; Donati, P.; Drazdauskas, A.; Chorniy, Y.; Bayo, A.; Kordopatis, G.

2017-11-01

To investigate the Galactic lithium enrichment history we se- lect well-measured main sequence field stars with UVES spectra from the GES iDR4 catalogue. In our selection, 1884 UVES stars are marked as field stars, including those of the Galactic disc and halo designated as MW (GEMW) fields, standard CoRoT (GES D_CR) field, standard radial velocity (GES DRV) field, and stars to the Galactic Bulge direction (GEMWBL). (1 data file).

Investigation on the radial micro-motion about piston of axial piston pump

Science.gov (United States)

Xu, Bing; Zhang, Junhui; Yang, Huayong; Zhang, Bin

2013-03-01

The limit working parameters and service life of axial piston pump are determined by the carrying ability and lubrication characteristic of its key friction pairs. Therefore, the design and optimization of the key friction pairs are always a key and difficult problem in the research on axial piston pump. In the traditional research on piston/cylinder pair, the assembly relationship of piston and cylinder bore is simplified into ideal cylindrical pair, which can not be used to analyze the influences of radial micro-motion of piston on the distribution characteristics of oil-film thickness and pressure in details. In this paper, based on the lubrication theory of the oil film, a numerical simulation model is built, taking the influences of roughness, elastic deformation of piston and pressure-viscosity effect into consideration. With the simulation model, the dynamic characteristics of the radial micro-motion and pressure distribution are analyzed, and the relationships between radial micro-motion and carrying ability, lubrication condition, and abrasion are discussed. Furthermore, a model pump for pressure distribution measurement of oil film between piston and cylinder bore is designed. The comparison of simulation and experimental results of pressure distribution shows that the simulation model has high accuracy. The experiment and simulation results demonstrate that the pressure distribution has peak values that are much higher than the boundary pressure in the piston chamber due to the radial micro-motion, and the abrasion of piston takes place mainly on the hand close to piston ball. In addition, improvement of manufacturing roundness and straightness of piston and cylinder bore is helpful to improve the carrying ability of piston/cylinder pair. The proposed research provides references for designing piston/cylinder pair, and helps to prolong the service life of axial piston pump.

Modeling of UF{sub 6} enrichment with gas centrifuges for nuclear safeguards activities

Energy Technology Data Exchange (ETDEWEB)

Mercurio, G.; Peerani, P.; Richir, P.; Janssens, W.; Eklund, G. [European Commission, Joint Research Centre, Institute for Transuranium Elements Via Fermi, 2749-TP181,20127 Ispra (Italy)

2012-09-26

The physical modeling of uranium isotopes ({sup 235}U, {sup 238}U) separation process by centrifugation of is a key aspect for predicting the nuclear fuel enrichment plant performances under surveillance by the Nuclear Safeguards Authorities. In this paper are illustrated some aspects of the modeling of fast centrifuges for UF{sub 6} gas enrichment and of a typical cascade enrichment plant with the Theoretical Centrifuge and Cascade Simulator (TCCS). The background theory for reproducing the flow field characteristics of a centrifuge is derived from the work of Cohen where the separation parameters are calculated using the solution of a differential enrichment equation. In our case we chose to solve the hydrodynamic equations for the motion of a compressible fluid in a centrifugal field using the Berman - Olander vertical velocity radial distribution and the solution was obtained using the Matlab software tool. The importance of a correct estimation of the centrifuge separation parameters at different flow regimes, lies in the possibility to estimate in a reliable way the U enrichment plant performances, once the separation external parameters are set (feed flow rate and feed, product and tails assays). Using the separation parameters of a single centrifuge allow to determine the performances of an entire cascade and, for this purpose; the software Simulink was used. The outputs of the calculation are the concentrations (assays) and the flow rates of the enriched (product) and depleted (tails) gas mixture. These models represent a valid additional tool, in order to verify the compliance of the U enrichment plant operator declarations with the 'on site' inspectors' measurements.

Status of Conceptual Design Progress for ITER Sector Sub-assembly Tools

Energy Technology Data Exchange (ETDEWEB)

Nam, Kyoung O; Park, Hyun Ki; Kim, Dong Jin [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Jae Hyuk; Kim, Kyung Kyu [SFA Engineering Corp., Changwon (Korea, Republic of); Im, Ki Hak; Robert, Shaw [ITER Organization, Paul lez Durance (France)

2010-05-15

-pit assembly tools that include lifting tools, radial beams, central column and supports

Influence of the radial spacing between cathodes on the surface composition of iron samples sintered by hollow cathode electric discharge

Directory of Open Access Journals (Sweden)

Brunatto S.F.

2001-01-01

Full Text Available The present work reports an investigation of the influence of the radial spacing between cathodes on the iron sintering process by hollow cathode electrical discharge, with surface enrichment of the alloying elements Cr and Ni. Pressed cylindrical samples of 9.5 mm diameter and density of 7.0 ± 0.1 g/cm³ were prepared by compaction of Ancorsteel 1000C iron powder. These samples, constituting the central cathode, were positioned concentrically in the interior of an external cathode machined from a tube of stainless steel AISI 310 (containing: 25% Cr, 16% Ni, 1.5% Mn, 1.5% Si, 0.03% C and the remainder Fe. Sintering was done at 1150 °C, for 120 min, utilizing radial spacings between the central and hollow cathodes of 3, 6 and 9 mm and a gas mixture of 80% Ar and 20% H2, with a flow rate of 5 cm³/s at a pressure of 3 Torr. The electric discharge was generated using a pulsed voltage power source, with a period of 200 mus. The radial spacing had only a slight influence on the quantity of atoms of alloying elements deposited and diffused on the surface of the sample. Analysis with a microprobe showed the presence of chrome (up to 4.0% and nickel (up to 3.0%, in at. % at the surface of the samples. This surface enrichment can be attributed to the mechanism of sputtering of the metallic atoms present in the external cathode, with the deposition of these elements on the sample surface and consequent diffusion within the sample.

Centrifugation. A theoretical study of oxygen enrichment by centrifugation

Energy Technology Data Exchange (ETDEWEB)

Kierkegaard, P.; Raetz, E.

1998-12-01

In the present paper we first investigate what happens if we fill a cylinder with air, close it and rotate it. The results show that no matter which peripheral speed is used, it is not possible by means of the radial separation effect alone, to enrich the oxygen concentration from the previous 21% to more then 23.3%, which is of no practical value. In case of a too low enrichment in one centrifuge, the wanted material from this centrifuge can be used as an input for a second centrifuge and so on, in this way forming a cascade of centrifuges. Oxygen will be enriched in each step, until the desired concentration is reached. Cascading was the technology in the very beginning by enrichment plants for uraniumhexaflouride, used for atomic weapons and nuclear power plants. In this study we try to avoid cascading by aiming for higher separation factors. Therefore, we next investigate the possibilities of using a countercurrent centrifuge where in principle the enriched gas is subjected to several centrifugation in the same centrifuge. The calculations show, that in this way it is possible to produce nearly a 100% pure oxygen (polluted with some heavier molecules like argon) in one machine. Our third step was to calculate the amount of oxygen produced per hour. Using a countercurrent centrifuge of the Zippe type, 100 cm high and 20 cm in diameter, it is or will be possible in the near future to produce 17 g enriched air per hour enriched to 50% oxygen. That corresponds to processing 1 m{sup 3} atmospherical air in the period of approximately 24 hours. This is not very impressive. Our fourth step was to estimate the amount of power used for producing this amount of oxygen. A rough, but complicated, estimate shows that the power consumption at the production level will be about the double of the consumption used today. The overall conclusion is, that centrifugation as a production method for oxygen (or nitrogen) will not be competitive with the currently used method in the

How enrichment affects exploration trade-offs in rats: implications for welfare and well-being.

Directory of Open Access Journals (Sweden)

Becca Franks

Full Text Available We propose that a comparative approach to well-being could be the key to understanding 'the good life.' Inspired by current theories of human well-being and animal welfare, we designed a novel test of exploration behavior. Environmentally and socially enriched Long-Evans female rats (N = 60 were trained in four simultaneously presented arms of an eight-arm radial-maze. They learned to expect successes in two arms and failures in the other two. After training, 20 animals remained in enriched housing (enrichment-maintenance while 40 animals were re-housed in standard, isolated conditions (enrichment-removal. Two weeks later, all animals were re-tested in the maze, initially with access to the four familiar arms only. In the final minute, they also had access to the unfamiliar ambiguous-arms. Though both groups showed significant interest in the ambiguous-arms (P.3. Thus, we show not only that rats will abandon known rewards and incur risk in order to explore, indicating that exploration is valuable in its own right, but also that individuals with (vs. without enriched housing conditions are more likely to engage in such exploratory behavior. This novel test contributes to the body of knowledge examining the importance of exploration in humans and other animals; implications for animal welfare and human well-being are discussed.

Perceived radial translation during centrifugation

NARCIS (Netherlands)

Bos, J.E.; Correia Grácio, B.J.

2015-01-01

BACKGROUND: Linear acceleration generally gives rise to translation perception. Centripetal acceleration during centrifugation, however, has never been reported giving rise to a radial, inward translation perception. OBJECTIVE: To study whether centrifugation can induce a radial translation

Evaluation of Spur Gear Pair on Tooth Root Bending Stress in Radial Misalignment Contact Condition

Directory of Open Access Journals (Sweden)

Lias M.R.

2014-07-01

Full Text Available This paper evaluates the effects of radial misalignment contact on the tooth root bending stress values of spur gear pair during the gear meshing cycle. Radial misalignment (H is denoted as the deviation of the pinion nominal position with respect to the gear tooth along the pinion axis to the gear which happened from manufacturing assembly errors (AE. A model based on involute 3D parametric CAD geometry, of spur gear design ISO 6336:2006 is analysed with allowable AE values from minimum 10μm to maximum 40μm with Finite-Element Method (FEM model based methodology using a dynamics module from ANSYS. Main parameters of interest are the Tooth root bending stress (TRBS in H condition with AE along the critical region with respect to face width of pinion-gear section. A comparison between standard High point single tooth contact models (HPSTC to this model showed a good agreement that H with AE had great influence on TRBS as the values’ increase. Radial misalignment influence factor (RMIF was introduced as indication of TRBS values in consideration of H due to AE shows and inverted patterns higher for pinion, give a good justification that the pinion is weaker compared to the gear.

Self-consistent radial sheath

International Nuclear Information System (INIS)

Hazeltine, R.D.

1988-12-01

The boundary layer arising in the radial vicinity of a tokamak limiter is examined, with special reference to the TEXT tokamak. It is shown that sheath structure depends upon the self-consistent effects of ion guiding-center orbit modification, as well as the radial variation of E /times/ B-induced toroidal rotation. Reasonable agreement with experiment is obtained from an idealized model which, however simplified, preserves such self-consistent effects. It is argued that the radial sheath, which occurs whenever confining magnetic field-lines lie in the plasma boundary surface, is an object of some intrinsic interest. It differs from the more familiar axial sheath because magnetized charges respond very differently to parallel and perpendicular electric fields. 11 refs., 1 fig

Evaluation of the radial design of fuel cells in an operation cycle of a BWR reactor

International Nuclear Information System (INIS)

Gonzalez C, J.; Martin del Campo M, C.

2003-01-01

This work is continuation of one previous in the one that the application of the optimization technique called Tabu search to the radial design of fuel cells of boiling water reactors (BWR, Boiling Water Reactor) is presented. The objective function used in the optimization process only include neutron parameters (k-infinite and peak of radial power) considering the cell at infinite media. It was obtained to reduce the cell average enrichment completing the characteristics of reactivity of an original cell. The objective of the present work is to validate the objective function that was used for the radial design of the fuel cell (test cell), analyzing the operation of a one cycle of the reactor in which fuels have been fresh recharged that contain an axial area with the nuclear database of the cell designed instead of the original cell. For it is simulated it with Cm-Presto the cycle 10 of the reactor operation of the Unit 1 of the Nuclear Power station of Laguna Verde (U1-CNLV). For the cycle evaluation its were applied so much the simulation with the Haling strategy, as the simulation of the one cycle with control rod patterns and they were evaluated the energy generation and several power limits and reactivity that are used as design parameters in fuel reloads of BWR reactors. The results at level of an operation cycle of the reactor, show that the objective function used in the optimization and radial design of the cell is adequate and that it can induce to one good use of the fuel. (Author)

Test of GERDA Phase II detector assembly

Energy Technology Data Exchange (ETDEWEB)

Bode, Tobias; Gusev, Konstantin [Technische Universitaet Muenchen (Germany); Schwingenheuer, Bernhard; Wagner, Victoria [Max-Planck Institut fuer Kernphysik, Heidelberg (Germany); Collaboration: GERDA-Collaboration

2014-07-01

The GERDA experiment searches for the lepton number violating neutrinoless double beta decay (0νββ) of {sup 76}Ge. The experiment uses HPGe detectors enriched in {sup 76}Ge as source and detection material. In GERDA Phase I five BEGe detectors were operated successfully. These detectors are distinguished for improved energy resolution and enhanced pulse shape discrimination (PSD) against background events. In Phase II additional 25 BEGe detectors will be installed. New electronics and radio-pure low-mass holders were specially designed for Phase II. Prior to the installation in GERDA all BEGe detectors are tested in their final assembly in the LNGS underground laboratory. This talk presents the mechanics and performance of the GERDA Phase II detector assembly.

Ultrasonic calibration assembly

International Nuclear Information System (INIS)

1981-01-01

Ultrasonic transducers for in-service inspection of nuclear reactor vessels have several problems associated with them which this invention seeks to overcome. The first is that of calibration or referencing a zero start point for the vertical axis of transducer movement to locate a weld defect. The second is that of verifying the positioning (vertically or at a predetermined angle). Thirdly there is the problem of ascertaining the speed per unit distance in the operating medium of the transducer beam prior to the actual inspection. The apparatus described is a calibration assembly which includes a fixed, generally spherical body having a surface for reflecting an ultrasonic beam from one of the transducers which can be moved until the reflection from the spherical body is the highest amplitude return signal indicating radial alignment from the body. (U.K.)

Filter assembly for metallic and intermetallic tube filters

Science.gov (United States)

Alvin, Mary Anne; Lippert, Thomas E.; Bruck, Gerald J.; Smeltzer, Eugene E.

2001-01-01

A filter assembly (60) for holding a filter element (28) within a hot gas cleanup system pressure vessel is provided, containing: a filter housing (62), said filter housing having a certain axial length and having a peripheral sidewall, said sidewall defining an interior chamber (66); a one piece, all metal, fail-safe/regenerator device (68) within the interior chamber (66) of the filter housing (62) and/or extending beyond the axial length of the filter housing, said device containing an outward extending radial flange (71) within the filter housing for seating an essential seal (70), the device also having heat transfer media (72) disposed inside and screens (80) for particulate removal; one compliant gasket (70) positioned next to and above the outward extending radial flange of the fail-safe/regenerator device; and a porous metallic corrosion resistant superalloy type filter element body welded at the bottom of the metal fail-safe/regenerator device.

Radial pseudoaneurysm following diagnostic coronary angiography

Directory of Open Access Journals (Sweden)

Shankar Laudari

2015-06-01

Full Text Available The radial artery access has gained popularity as a method of diagnostic coronary catheterization compared to femoral artery puncture in terms of vascular complications and early ambulation. However, very rare complication like radial artery pseudoaneurysm may occur following cardiac catheterization which may give rise to serious consequences. Here, we report a patient with radial pseudoaneurysm following diagnostic coronary angiography. Adequate and correct methodology of compression of radial artery following puncture for maintaining hemostasis is the key to prevention.DOI: http://dx.doi.org/10.3126/jcmsn.v10i3.12776 Journal of College of Medical Sciences-Nepal, 2014, Vol-10, No-3, 48-50

Enrichment dynamics of Listeria monocytogenes and the associated microbiome from naturally contaminated ice cream linked to a listeriosis outbreak.

Science.gov (United States)

Ottesen, Andrea; Ramachandran, Padmini; Reed, Elizabeth; White, James R; Hasan, Nur; Subramanian, Poorani; Ryan, Gina; Jarvis, Karen; Grim, Christopher; Daquiqan, Ninalynn; Hanes, Darcy; Allard, Marc; Colwell, Rita; Brown, Eric; Chen, Yi

2016-11-16

Microbiota that co-enrich during efforts to recover pathogens from foodborne outbreaks interfere with efficient detection and recovery. Here, dynamics of co-enriching microbiota during recovery of Listeria monocytogenes from naturally contaminated ice cream samples linked to an outbreak are described for three different initial enrichment formulations used by the Food and Drug Administration (FDA), the International Organization of Standardization (ISO), and the United States Department of Agriculture (USDA). Enrichment cultures were analyzed using DNA extraction and sequencing from samples taken every 4 h throughout 48 h of enrichment. Resphera Insight and CosmosID analysis tools were employed for high-resolution profiling of 16S rRNA amplicons and whole genome shotgun data, respectively. During enrichment, other bacterial taxa were identified, including Anoxybacillus, Geobacillus, Serratia, Pseudomonas, Erwinia, and Streptococcus spp. Surprisingly, incidence of L. monocytogenes was proportionally greater at hour 0 than when tested 4, 8, and 12 h later with all three enrichment schemes. The corresponding increase in Anoxybacillus and Geobacillus spp.indicated these taxa co-enriched in competition with L. monocytogenes during early enrichment hours. L. monocytogenes became dominant after 24 h in all three enrichments. DNA sequences obtained from shotgun metagenomic data of Listeria monocytogenes at 48 h were assembled to produce a consensus draft genome which appeared to have a similar tracking utility to pure culture isolates of L. monocytogenes. All three methods performed equally well for enrichment of Listeria monocytogenes. The observation of potential competitive exclusion of L. mono by Anoxybacillus and Geobacillus in early enrichment hours provided novel information that may be used to further optimize enrichment formulations. Application of Resphera Insight for high-resolution analysis of 16S amplicon sequences accurately identified L. monocytogenes

Aptamer-assembled nanomaterials for fluorescent sensing and imaging

Science.gov (United States)

Lu, Danqing; He, Lei; Zhang, Ge; Lv, Aiping; Wang, Ruowen; Zhang, Xiaobing; Tan, Weihong

2017-01-01

Aptamers, which are selected in vitro by a technology known as the systematic evolution of ligands by exponential enrichment (SELEX), represent a crucial recognition element in molecular sensing. With advantages such as good biocompatibility, facile functionalization, and special optical and physical properties, various nanomaterials can protect aptamers from enzymatic degradation and nonspecific binding in living systems and thus provide a preeminent platform for biochemical applications. Coupling aptamers with various nanomaterials offers many opportunities for developing highly sensitive and selective sensing systems. Here, we focus on the recent applications of aptamer-assembled nanomaterials in fluorescent sensing and imaging. Different types of nanomaterials are examined along with their advantages and disadvantages. Finally, we look toward the future of aptamer-assembled nanomaterials.

Structure and experimental program for SNEAK 12 - an assembly for fast breeder safety experiments

International Nuclear Information System (INIS)

Helm, F.

1979-01-01

The critical assembly SNEAK 12 and its foreseen experimental program have the main purpose to check the validity of neutron physics calculational methods used in the analysis of accidental situations of fast breeder reactors. In the investigation of accidental courses configurations have to be considered, which are caused by assembly deformation and meltdown and which are characterized by irregular fuel and structural material arrangements with cavities and empty channels. The reactivity differences between the unperturbed core and a series of perturbed configurations have to be determined. The individual configurations have to be chosen in such a way, that the calculational methods for the different aspects of the accident sequence (formation of cavities and channels, relocation of fissile and fertile material and steel) can be tested one by one. Two different cores are foreseen: SNEAK 12A as a one-zone core with enriched uranium fuel and SNEAK 12B with a central test zone with plutonium-uranium mixed-oxide fuel surrounded by a driver zone of enriched uranium. The report describes these cores and their assemblies, and the experimental program is outlined

Endoscopic versus open radial artery harvest and mammario-radial versus aorto-radial grafting in patients undergoing coronary artery bypass surgery

DEFF Research Database (Denmark)

Carranza, Christian L; Ballegaard, Martin; Werner, Mads U

2014-01-01

the postoperative complications will be registered, and we will evaluate muscular function, scar appearance, vascular supply to the hand, and the graft patency including the patency of the central radial artery anastomosis. A patency evaluation by multi-slice computer tomography will be done at one year...... to aorto-radial revascularisation techniques but this objective is exploratory. TRIAL REGISTRATION: ClinicalTrials.gov identifier: NCT01848886.Danish Ethics committee number: H-3-2012-116.Danish Data Protection Agency: 2007-58-0015/jr.n:30-0838....

Fuel assembly and nuclear reactor core

International Nuclear Information System (INIS)

Masumi, Ryoji; Aoyama, Motoo; Yamashita, Jun-ichi.

1995-01-01

The present invention concerns a fuel assembly and a nuclear reactor core capable of improving a transmutation rate of transuranium elements while improving a residual rate of fission products. In a reactor core of a BWR type reactor to which fuel rods with transuranium elements (TRU) enriched are loaded, the enrichment degree of transuranium elements occupying in fuel materials is determined not less than 2wt%, as well as a ratio of number of atoms between hydrogen and fuel heavy metals in an average reactor core under usual operation state (H/HM) is determined not more than 3 times. In addition, a ratio of the volumes between coolant regions and fuel material regions is determined not more than 2 times. A T ratio (TRU/Pu) is lowered as the TRU enrichment degree is higher and the H/HM ratio is lower. In order to reduce the T ratio not more than 1, the TRU enrichment degree is determined as not less than 2wt%, and the H/HM ratio is determined to not more than 3 times. Accordingly, since the H/HM ratio is reduced to not more than 1, and TRU is transmuted while recycling it with plutonium, the transmutation ratio of transuranium elements can be improved while improving the residual rate of fission products. (N.H.)

ZOMG - III. The effect of halo assembly on the satellite population

Science.gov (United States)

Garaldi, Enrico; Romano-Díaz, Emilio; Borzyszkowski, Mikolaj; Porciani, Cristiano

2018-01-01

We use zoom hydrodynamical simulations to investigate the properties of satellites within galaxy-sized dark-matter haloes with different assembly histories. We consider two classes of haloes at redshift z = 0: 'stalled' haloes that assembled at z > 1 and 'accreting' ones that are still forming nowadays. Previously, we showed that the stalled haloes are embedded within thick filaments of the cosmic web, while the accreting ones lie where multiple thin filaments converge. We find that satellites in the two classes have both similar and different properties. Their mass spectra, radial count profiles, baryonic and stellar content, and the amount of material they shed are indistinguishable. However, the mass fraction locked in satellites is substantially larger for the accreting haloes as they experience more mergers at late times. The largest difference is found in the satellite kinematics. Substructures fall towards the accreting haloes along quasi-radial trajectories whereas an important tangential velocity component is developed, before accretion, while orbiting the filament that surrounds the stalled haloes. Thus, the velocity anisotropy parameter of the satellites (β) is positive for the accreting haloes and negative for the stalled ones. This signature enables us to tentatively categorize the Milky Way halo as stalled based on a recent measurement of β. Half of our haloes contain clusters of satellites with aligned orbital angular momenta corresponding to flattened structures in space. These features are not driven by baryonic physics and are only found in haloes hosting grand-design spiral galaxies, independently of their assembly history.

Numerical simulation of liquid-metal-flows in radial-toroidal-radial bends

International Nuclear Information System (INIS)

Molokov, S.; Buehler, L.

1993-09-01

Magnetohydrodynamic flows in a U-bend and right-angle bend are considered with reference to the radial-toroidal-radial concept of a self-cooled liquid-metal blanket. The ducts composing bends have rectangular cross-section. The applied magnetic field is aligned with the toroidal duct and perpendicular to the radial ones. At high Hartmann number the flow region is divided into cores and boundary layers of different types. The magnetohydrodynamic equations are reduced to a system of partial differential equations governing wall electric potentials and the core pressure. The system is solved numerically by two different methods. The first method is iterative with iteration between wall potential and the core pressure. The second method is a general one for the solution of the core flow equations in curvilinear coordinates generated by channel geometry and magnetic field orientation. Results obtained are in good agreement. They show, that the 3D-pressure drop of MHD flows in a U-bend is not a critical issue for blanket applications. (orig./HP) [de

Development of IAEA safeguards at low enrichment uranium fuel fabrication plants

International Nuclear Information System (INIS)

Badawy, I.

1988-01-01

In this report the nuclear material at low enrichment uranium fuel fabrication plants under IAEA safeguards is studied. The current verification practices of the nuclear material and future improvements are also considered. The problems met during the implementation of the the verification measures of the nuclear material - particularly for the fuel assemblies are discussed. The additional verification activities as proposed for future improvements are also discussed including the physical inventory verification and the verification of receipts and shipments. It is concluded that the future development of the present IAEA verification practices at low enrichment uranium fuel fabrication plants would necessitate the application of quantitative measures of the nuclear material and the implementation of advanced measurement techniques and instruments. 2 fig., 4 tab

Nonlinear Modeling by Assembling Piecewise Linear Models

Science.gov (United States)

Yao, Weigang; Liou, Meng-Sing

2013-01-01

To preserve nonlinearity of a full order system over a parameters range of interest, we propose a simple modeling approach by assembling a set of piecewise local solutions, including the first-order Taylor series terms expanded about some sampling states. The work by Rewienski and White inspired our use of piecewise linear local solutions. The assembly of these local approximations is accomplished by assigning nonlinear weights, through radial basis functions in this study. The efficacy of the proposed procedure is validated for a two-dimensional airfoil moving at different Mach numbers and pitching motions, under which the flow exhibits prominent nonlinear behaviors. All results confirm that our nonlinear model is accurate and stable for predicting not only aerodynamic forces but also detailed flowfields. Moreover, the model is robustness-accurate for inputs considerably different from the base trajectory in form and magnitude. This modeling preserves nonlinearity of the problems considered in a rather simple and accurate manner.

Plutonium assemblies in reload 1 of the Dodewaard Reactor

International Nuclear Information System (INIS)

Bairiot, H.; Deramaix, P.; Vandenberg, C.; Leenders, L.; Mostert, P.

1977-01-01

Since 1963, Belgonucleaire has been developing the design of plutonium assemblies of the island type (i.e., plutonium rods inserted in the control zone of the assembly and enriched uranium rods at the periphery) for light water reactors. The application to boiling water reactors (BWRs) led to the introduction, in April 1971, of two prototype plutonium island assemblies in the Dodewaard BWR (The Netherlands): Those assemblies incorporating plutonium in 42 percent of the rods are interchangeable with standard uranium assemblies of the same reload. Their design, which had to meet these criteria, was performed using the routine order in use at Belgonucleaire; experimental checks included a mock-up configuration simulated in the VENUS critical facility at Mol and open-vessel cold critical experiments performed in the Dodewaard core. The pelleted plutonium rods were fabricated and controlled by Belgonucleaire following the manufacturing procedures developed at the production plant. In one of the assemblies, three vibrated plutonium fuel rods with a lower fuel density were introduced in the three most highly rated positions to reduce the power rating. Those plutonium assemblies experienced peak pellet ratings up to 535 W/cm and were discharged in April 1974 after having reached a mean burnup of approximately 21,000 MWd/MT. In-core instrumentation during operation, visual examinations, and reactivity substitution experiments during reactor shutdown did not indicate any special feature for those assemblies compared to the standard uranium assemblies, thereby demonstrating their interchangeability

Self-assembled Thiolated Calix[n]arene (n=4, 6, 8) Films on Gold Electrodes and Application for Electrochemical Determination Dopamine

International Nuclear Information System (INIS)

Zheng, Gang; Chen, Ming; Liu, Xinyue; Zhou, Jun; Xie, Ju; Diao, Guowang

2014-01-01

Highlights: • TCnA/GE was prepared by using a simple self-assembled strategy. • Multilayer self-assembled films of TCnA molecules were fabricated on GE. • TCnA/GE exhibited high supramolecular recognition and enrichment capability. • TC8A/GE showed excellent electrochemical performance for DA. - Abstract: In this study, gold electrodes (GE) modified with three kinds of thiolated calix[4,6,8]arenes (TCnA: TC4A, TC6A, TC8A) were successfully prepared using a simple self-assembly strategy. Three self-assembled films were characterized by cyclic voltammetry measurement, electrochemical impedance spectroscopy, static contact angle measurement and atomic force microscopy. The results confirmed that TCnA molecules effectively absorbed onto the surface of gold electrodes to fabricate the multilayer self-assembled films. Cyclic voltammetry (CV) and differential pulse voltammetry (DPV) measurement showed that the TCnA/GE exhibited high supramolecular recognition and enrichment capability and consequently displayed good electrochemical response toward dopamine (DA). Especially, TC8A/GE exhibited an excellent electrochemical performance for DA with high current densities of 1.5 mA mmol −1 L cm −2 , broad linear range (1 × 10 −6 to 1 × 10 −3 mol L −1 ) and low detection limit (5 × 10 −7 mol L −1 ). The mechanism of supramolecular recognition and enrichment capability of TCnA/GE was discussed

Dedicated radial ventriculography pigtail catheter

Energy Technology Data Exchange (ETDEWEB)

Vidovich, Mladen I., E-mail: miv@uic.edu

2013-05-15

A new dedicated cardiac ventriculography catheter was specifically designed for radial and upper arm arterial access approach. Two catheter configurations have been developed to facilitate retrograde crossing of the aortic valve and to conform to various subclavian, ascending aortic and left ventricular anatomies. The “short” dedicated radial ventriculography catheter is suited for horizontal ascending aortas, obese body habitus, short stature and small ventricular cavities. The “long” dedicated radial ventriculography catheter is suited for vertical ascending aortas, thin body habitus, tall stature and larger ventricular cavities. This new design allows for improved performance, faster and simpler insertion in the left ventricle which can reduce procedure time, radiation exposure and propensity for radial artery spasm due to excessive catheter manipulation. Two different catheter configurations allow for optimal catheter selection in a broad range of patient anatomies. The catheter is exceptionally stable during contrast power injection and provides equivalent cavity opacification to traditional femoral ventriculography catheter designs.

Nondestructive Assay Data Integration with the SKB-50 Assemblies - FY16 Update

International Nuclear Information System (INIS)

Tobin, Stephen Joseph; Fugate, Michael Lynn; Trellue, Holly Renee; DeBaere, Paul; Sjoland, Anders; Liljenfeldt, Henrik; Hu, Jianwei; Backstrom, Ulrika; Bengtsson, Martin; Burr, Tomas; Eliasson, Annika; Favalli, Andrea; Gauld, Ian; Grogan, Brandon; Jansson, Peter; Junell, Henrik; Schwalbach, Peter; Vaccaro, Stefano; Vo, Duc Ta; Wildestrand, Henrik

2016-01-01

A project to research the application of non-destructive assay (NDA) techniques for spent fuel assemblies is underway at the Central Interim Storage Facility for Spent Nuclear Fuel (for which the Swedish acronym is Clab) in Oskarshamn, Sweden. The research goals of this project contain both safeguards and non-safeguards interests. These nondestructive assay (NDA) technologies are designed to strengthen the technical toolkit of safeguard inspectors and others to determine the following technical goals more accurately; Verify initial enrichment, burnup, and cooling time of facility declaration for spent fuel assemblies; Detect replaced or missing pins from a given spent fuel assembly to confirm its integrity; and Estimate plutonium mass and related plutonium and uranium fissile mass parameters in spent fuel assemblies. Estimate heat content, and measure reactivity (multiplication).

Nondestructive Assay Data Integration with the SKB-50 Assemblies - FY16 Update

Energy Technology Data Exchange (ETDEWEB)

Tobin, Stephen Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fugate, Michael Lynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Trellue, Holly Renee [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); DeBaere, Paul [DG Energy, Luxembourg (Germany); Sjoland, Anders [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Liljenfeldt, Henrik [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hu, Jianwei [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Backstrom, Ulrika [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Vattenfall AB, Stockholm (Sweden); Bengtsson, Martin [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Vattenfall AB, Stockholm (Sweden); Burr, Tomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); International Atomic Energy Agency, Vienna (Austria); Eliasson, Annika [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Favalli, Andrea [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Gauld, Ian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Grogan, Brandon [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jansson, Peter [Uppsala Univ. (Sweden); Junell, Henrik [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Schwalbach, Peter [DG Energy, Luxembourg (Germany); Vaccaro, Stefano [DG Energy, Luxembourg (Germany); Vo, Duc Ta [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wildestrand, Henrik [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Vattenfall AB, Stockholm (Sweden)

2016-10-28

A project to research the application of non-destructive assay (NDA) techniques for spent fuel assemblies is underway at the Central Interim Storage Facility for Spent Nuclear Fuel (for which the Swedish acronym is Clab) in Oskarshamn, Sweden. The research goals of this project contain both safeguards and non-safeguards interests. These nondestructive assay (NDA) technologies are designed to strengthen the technical toolkit of safeguard inspectors and others to determine the following technical goals more accurately; Verify initial enrichment, burnup, and cooling time of facility declaration for spent fuel assemblies; Detect replaced or missing pins from a given spent fuel assembly to confirm its integrity; and Estimate plutonium mass and related plutonium and uranium fissile mass parameters in spent fuel assemblies. Estimate heat content, and measure reactivity (multiplication).

Radial wave crystals: radially periodic structures from anisotropic metamaterials for engineering acoustic or electromagnetic waves.

Science.gov (United States)

Torrent, Daniel; Sánchez-Dehesa, José

2009-08-07

We demonstrate that metamaterials with anisotropic properties can be used to develop a new class of periodic structures that has been named radial wave crystals. They can be sonic or photonic, and wave propagation along the radial directions is obtained through Bloch states like in usual sonic or photonic crystals. The band structure of the proposed structures can be tailored in a large amount to get exciting novel wave phenomena. For example, it is shown that acoustical cavities based on radial sonic crystals can be employed as passive devices for beam forming or dynamically orientated antennas for sound localization.

Development of a New Multiplying Assembly for Research, Validation, Evaluation, and Learning

Energy Technology Data Exchange (ETDEWEB)

David L. Chichester

2012-10-01

A new multiplying test assembly is under development at Idaho National Laboratory (INL) to support research, validation, evaluation, and learning. The item is comprised of two stacked highly-enriched uranium (HEU) cylinders each 11.4 cm in diameter and having a combined height of 8.4 cm. The combined mass is 14.4 kg of HEU. Calculations for the bare configuration of the assembly indicate a multiplication level of >2.5 (keff = 0.62). Reflected configurations of the assembly, using either polyethylene or tungsten, are possible and have the capability of raising its multiplication level to approximately 8. This paper will describe the MCNP calculations performed to assess the assembly's multiplication level under different conditions and describe the resource available at INL to support visiting researchers in their use of the material. We will also describe some preliminary calculations and test activities using the assembly to study neutron multiplicity.

An investigative approach to explore optimum assembly process design for annular targets carrying LEU foil

Science.gov (United States)

Hoyer, Annemarie

Technetium-99m is the most widely used nuclear isotope in the medical field, with nearly 80 to 85% of all diagnostic imaging procedures. The daughter isotope of molybdenum-99 is currently produced using weapons-grade uranium. A suggested design for aluminum targets carrying low-enriched uranium (LEU) foil is presented for the fulfillment of eliminating highly enriched uranium (HEU) for medical isotope production. The assembly process that this research focuses on is the conventional draw-plug process which is currently used and lastly the sealing process. The research is unique in that it is a systematic approach to explore the optimal target assembly process to produce those targets with the required quality and integrity. Conducting 9 parametric experiments, aluminum tubes with a nickel foil fission-barrier and a surrogate stainless steel foil are assembled, welded and then examined to find defects, to determine residual stresses, and to find the best cost-effective target dimensions. The experimental design consists of 9 assembly combinations that were found through orthogonal arrays in order to explore the significance of each factor. Using probabilistic modeling, the parametric study is investigated using the Taguchi method of robust analysis. Depending on the situation, optimal conditions may be a nominal, a minimized or occasionally a maximized condition. The results will provide the best target design and will give optimal quality with little or no assembly defects.

The necessity of improvement for the current LWR fuel assembly homogenization method

International Nuclear Information System (INIS)

Tang Chuntao; Huang Hao; Zhang Shaohong

2007-01-01

When the modern LWR core analysis method is used to do core nuclear design and in-core fuel management calculation, how to accurately obtain the fuel assembly homogenized parameters is a crucial issue. In this paper, taking the NEA C5G7-MOX benchmark problem as a severe test problem, which involves low-enriched uranium assemblies interspersed with MOX assemblies, we have re-examined the applicability of the two major assumptions of the modern equivalence theory for fuel assembly homoge- nization, i.e. the isolated assembly spatial spectrum assumption and the condensed two- group representation assumption. Numerical results have demonstrated that for LWR cores with strong spectrum interaction, both of these two assumptions are no longer applicable and the improvement for the homogenization method is necessary, the current two-group representation should be improved by the multigroup representation and the current reflective assembly boundary condition should be improved by the 'real' assembly boundary condition. This is a research project supported by National Natural Science Foundation of China (10605016). (authors)

A KiDS weak lensing analysis of assembly bias in GAMA galaxy groups

Science.gov (United States)

Dvornik, Andrej; Cacciato, Marcello; Kuijken, Konrad; Viola, Massimo; Hoekstra, Henk; Nakajima, Reiko; van Uitert, Edo; Brouwer, Margot; Choi, Ami; Erben, Thomas; Fenech Conti, Ian; Farrow, Daniel J.; Herbonnet, Ricardo; Heymans, Catherine; Hildebrandt, Hendrik; Hopkins, Andrew M.; McFarland, John; Norberg, Peder; Schneider, Peter; Sifón, Cristóbal; Valentijn, Edwin; Wang, Lingyu

2017-07-01

We investigate possible signatures of halo assembly bias for spectroscopically selected galaxy groups from the Galaxy And Mass Assembly (GAMA) survey using weak lensing measurements from the spatially overlapping regions of the deeper, high-imaging-quality photometric Kilo-Degree Survey. We use GAMA groups with an apparent richness larger than 4 to identify samples with comparable mean host halo masses but with a different radial distribution of satellite galaxies, which is a proxy for the formation time of the haloes. We measure the weak lensing signal for groups with a steeper than average and with a shallower than average satellite distribution and find no sign of halo assembly bias, with the bias ratio of 0.85^{+0.37}_{-0.25}, which is consistent with the Λ cold dark matter prediction. Our galaxy groups have typical masses of 1013 M⊙ h-1, naturally complementing previous studies of halo assembly bias on galaxy cluster scales.

Aneurisma idiopático de artéria radial: relato de caso Idiopathic radial artery aneurysm: case report

Directory of Open Access Journals (Sweden)

Luiz Ernani Meira Jr.

2011-12-01

Full Text Available Os aneurismas da artéria radial são extremamente raros. Em sua maioria, consistem de pseudoaneurismas pós-traumáticos. Os aneurismas da artéria radial verdadeiros podem ser idiopáticos, congênitos, pós-estenóticos ou associados a patologias, tais como vasculites e doenças do tecido conjuntivo. Foi relatado um caso de aneurisma idiopático de artéria radial em uma criança de três anos, que, após completa investigação diagnóstica complementar, foi submetida à ressecção cirúrgica.Radial artery aneurysms are extremely rare. Post-traumatic pseudoaneurysms are the vast majority. True radial artery aneurysms can be idiopathic, congenital, poststenotic, or associated with some pathologies, such as vasculitis and conjunctive tissue diseases. We report a case of an idiopathic aneurysm of the radial artery in a three-year-old child who was submitted to surgical resection after a complete diagnostic approach.

Analysis of a PHWR slightly enriched fuel

International Nuclear Information System (INIS)

Notari, C.; Marajofsky, A.

1994-01-01

It is widely known that the use of slightly enriched uranium in PHWR reactors presents economic advantages derived from the fact that less uranium is required for producing the same amount of energy. Several studies related with the use of this alternative in Atucha I NPP have been performed. The fuel assembly geometry considered up to now has been almost identical to the natural uranium one. In this work a modification consisting in the use of annular pellets in the outer ring of the cluster is analyzed. This design produces several performance benefits. The redistribution of the power in the fuel improves the maximum to average bundle power ratio. The improvement achieved depends on the void volume in the pellets which at the same time represents a certain burnup decrease. These parameters (power ratios and burnup loss) are quantified for the Atucha I and Embalse NPPs. This design improves the fuel behaviour with respect to the burnup extension derived from the slight enrichment. It is also interesting in case an overall power increase is considered. (author). 16 refs, 8 figs, 1 tab

Ulnar nerve entrapment complicating radial head excision

Directory of Open Access Journals (Sweden)

Kevin Parfait Bienvenu Bouhelo-Pam

Full Text Available Introduction: Several mechanisms are involved in ischemia or mechanical compression of ulnar nerve at the elbow. Presentation of case: We hereby present the case of a road accident victim, who received a radial head excision for an isolated fracture of the radial head and complicated by onset of cubital tunnel syndrome. This outcome could be the consequence of an iatrogenic valgus of the elbow due to excision of the radial head. Hitherto the surgical treatment of choice it is gradually been abandoned due to development of radial head implant arthroplasty. However, this management option is still being performed in some rural centers with low resources. Discussion: The radial head plays an important role in the stability of the elbow and his iatrogenic deformity can be complicated by cubital tunnel syndrome. Conclusion: An ulnar nerve release was performed with favorable outcome. Keywords: Cubital tunnel syndrome, Peripheral nerve palsy, Radial head excision, Elbow valgus

Solution High-Energy Burst Assembly (SHEBA) results from subprompt critical experiments with uranyl fluoride fuel

International Nuclear Information System (INIS)

Cappiello, C.C.; Butterfield, K.B.; Sanchez, R.G.

1997-10-01

The Solution High-Energy Burst Assembly (SHEBA) was originally constructed during 1980 and was designed to be a clean free-field geometry, right-circular, cylindrically symmetric critical assembly employing U(5%)O 2 F 2 solution as fuel. A second version of SHEBA, employing the same fuel but equipped with a fuel pump and shielding pit, was commissioned in 1993. This report includes data and operating experience for the 1993 SHEBA only. Solution-fueled benchmark work focused on the development of experimental measurements of the characterization of SHEBA; a summary of the results are given. A description of the system and the experimental results are given in some detail in the report. Experiments were designed to: (1) study the behavior of nuclear excursions in a low-enrichment solution, (2) evaluate accidental criticality alarm detectors for fuel-processing facilities, (3) provide radiation spectra and dose measurements to benchmark radiation transport calculations on a low-enrichment solution system similar to centrifuge enrichment plants, and (4) provide radiation fields to calibrate personnel dosimetry. 15 refs., 37 figs., 10 tabs

Bouyancy effects on sodium coolant temperature profiles measured in an electrically heated mock-up of a 61-rod breeder reactor blanket assembly

International Nuclear Information System (INIS)

Engel, F.C.; Markley, R.A.; Minushkin, B.

1978-01-01

The paper describes test results selected to demonstrate the effect of buoyancy on the temperature profiles in a 61-rod electrically heated mock-up of an LMFBR radial blanket assembly. In these assemblies, heat transfer occurs over a wide range of complex operating conditions. The range and complexity of conditions are the result of the steep flux and power gradients which are an inherent feature of the blanket region and the power generation level in an assembly which can vary from 20 to 1100 kW

Refinement of criticality and breeding parameters by means of experiments on a series of critical assemblies

International Nuclear Information System (INIS)

Golubev, V.I.; Dulin, V.A.; Kazanskij, Yu.A.; Mamontov, V.M.; Mozhaev, V.K.; Sidorov, G.I.

1980-01-01

A programme of measurements was performed on a number of critical assemblies with the aim of obtaining reliable experimental data under conditions approximating the simplest calculation model. To this end the neutron balance at the centres of the BFS-31, BFS-33, BFS-35, BFS-38, KBR-3 and KBR-7 critical assemblies was investigated. These assemblies contained central inserts made of uranium dioxide (BFS-33), natural uranium oxide and plutonium metal (BFS-31), natural uranium and plutonium metal (BFS-38), 90% enriched metallic uranium and stainless steel (KBR-3) and enriched uranium dioxide and nickel (KBR-7). The composition of the inserts was such that Ksub(infinite)=1. The K + values, the ratios of the reaction rates of the principal raw material and fissionable isotopes and the reactivity coefficients of a number of materials were measured in the inserts. The components of the breeding coefficient were measured at the centre of the BFS-39 critical assembly which simulates a power reactor (simplest composition with low- and high-enrichment zones and no control mechanism). The authors describe briefly the critical assemblies, the methods of measurement and calculation and methods of correcting for differences between the calculation model and the conditions under which the measurements were performed and compare the results of the experiments with the corresponding theoretical values obtained using various systems of group constants. In their latest versions, the group constants derived from different sets of integral experiments describe the experimental data much better than was previously possible. The deviations which occur in the predicted criticality and breeding parameters using different versions of the constants essentially reflect the difference in the results of the sets of integral experiments that were used for the group constants. (author)

First assembly phase for the ATLAS toroid coils

CERN Multimedia

Maximilien Brice

2003-01-01

The ATLAS barrel toroid system consists of eight coils, each of axial length 25.3 m, assembled radially and symmetrically around the beam axis. The coils are of a flat racetrack type with two double-pancake windings made of 20.5 kA aluminium-stabilized niobium-titanium superconductor. In the first phase of assembly, the two 'pancakes' are packed into their vacuum vessel. This is done using bladders filled with resin and glass microbeads under pressure. The resin is heated and, once cooled, holds the pancakes in place. The operation has to be performed on both sides of the coil, which necessitated a special technique to turn the coils over and then transport them to the heating table. Photos 01, 02, 03: Transporting the coil to the heating table using a special lifting gantry manufactured at JINR-Dubna, Russia in preparation for the 'bladderisation' operation.

Radial lean direct injection burner

Science.gov (United States)

Khan, Abdul Rafey; Kraemer, Gilbert Otto; Stevenson, Christian Xavier

2012-09-04

A burner for use in a gas turbine engine includes a burner tube having an inlet end and an outlet end; a plurality of air passages extending axially in the burner tube configured to convey air flows from the inlet end to the outlet end; a plurality of fuel passages extending axially along the burner tube and spaced around the plurality of air passage configured to convey fuel from the inlet end to the outlet end; and a radial air swirler provided at the outlet end configured to direct the air flows radially toward the outlet end and impart swirl to the air flows. The radial air swirler includes a plurality of vanes to direct and swirl the air flows and an end plate. The end plate includes a plurality of fuel injection holes to inject the fuel radially into the swirling air flows. A method of mixing air and fuel in a burner of a gas turbine is also provided. The burner includes a burner tube including an inlet end, an outlet end, a plurality of axial air passages, and a plurality of axial fuel passages. The method includes introducing an air flow into the air passages at the inlet end; introducing a fuel into fuel passages; swirling the air flow at the outlet end; and radially injecting the fuel into the swirling air flow.

Modular enrichment measurement system for in-situ enrichment assay

International Nuclear Information System (INIS)

Stewart, J.P.

1976-01-01

A modular enrichment measurement system has been designed and is in operation within General Electric's Nuclear Fuel Fabrication Facility for the in-situ enrichment assay of uranium-bearing materials in process containers. This enrichment assay system, which is based on the ''enrichment meter'' concept, is an integral part of the site's enrichment control program and is used in the in-situ assay of the enrichment of uranium dioxide (UO 2 ) powder in process containers (five gallon pails). The assay system utilizes a commercially available modular counting system and a collimnator designed for compatability with process container transport lines and ease of operator access. The system has been upgraded to include a microprocessor-based controller to perform system operation functions and to provide data acquisition and processing functions. Standards have been fabricated and qualified for the enrichment assay of several types of uranium-bearing materials, including UO 2 powders. The assay system has performed in excess of 20,000 enrichment verification measurements annually and has significantly contributed to the facility's enrichment control program

Anomalies of radial and ulnar arteries

Directory of Open Access Journals (Sweden)

Rajani Singh

Full Text Available Abstract During dissection conducted in an anatomy department of the right upper limb of the cadaver of a 70-year-old male, both origin and course of the radial and ulnar arteries were found to be anomalous. After descending 5.5 cm from the lower border of the teres major, the brachial artery anomalously bifurcated into a radial artery medially and an ulnar artery laterally. In the arm, the ulnar artery lay lateral to the median nerve. It followed a normal course in the forearm. The radial artery was medial to the median nerve in the arm and then, at the level of the medial epicondyle, it crossed from the medial to the lateral side of the forearm, superficial to the flexor muscles. The course of the radial artery was superficial and tortuous throughout the arm and forearm. The variations of radial and ulnar arteries described above were associated with anomalous formation and course of the median nerve in the arm. Knowledge of neurovascular anomalies are important for vascular surgeons and radiologists.

Variations in the usage and composition of a radial cocktail during radial access coronary angiography procedures.

LENUS (Irish Health Repository)

Pate, G

2011-10-01

A survey was conducted of medication administered during radial artery cannulation for coronary angiography in 2009 in Ireland; responses were obtained for 15 of 20 centres, in 5 of which no radial access procedures were undertaken. All 10 (100%) centres which provided data used heparin and one or more anti-spasmodics; verapamil in 9 (90%), nitrate in 1 (10%), both in 2 (20%). There were significant variations in the doses used. Further work needs to be done to determine the optimum cocktail to prevent radial artery injury following coronary angiography.

Validation of Noninvasive MOEMS-Assisted Measurement System Based on CCD Sensor for Radial Pulse Analysis

Directory of Open Access Journals (Sweden)

Rolanas Dauksevicius

2013-04-01

Full Text Available Examination of wrist radial pulse is a noninvasive diagnostic method, which occupies a very important position in Traditional Chinese Medicine. It is based on manual palpation and therefore relies largely on the practitioner’s subjective technical skills and judgment. Consequently, it lacks reliability and consistency, which limits practical applications in clinical medicine. Thus, quantifiable characterization of the wrist pulse diagnosis method is a prerequisite for its further development and widespread use. This paper reports application of a noninvasive CCD sensor-based hybrid measurement system for radial pulse signal analysis. First, artery wall deformations caused by the blood flow are calibrated with a laser triangulation displacement sensor, following by the measurement of the deformations with projection moiré method. Different input pressures and fluids of various viscosities are used in the assembled artificial blood flow system in order to test the performance of laser triangulation technique with detection sensitivity enhancement through microfabricated retroreflective optical element placed on a synthetic vascular graft. Subsequently, the applicability of double-exposure whole-field projection moiré technique for registration of blood flow pulses is considered: a computational model and representative example are provided, followed by in vitro experiment performed on a vascular graft with artificial skin atop, which validates the suitability of the technique for characterization of skin surface deformations caused by the radial pulsation.

Design of radial reinforcement for prestressed concrete containments

Energy Technology Data Exchange (ETDEWEB)

Wang, Shen, E-mail: swang@bechtel.com [Bechtel Power Corporation, 5275 Westview Drive, BP2-2C3, Frederick, MD 21703 (United States); Munshi, Javeed A., E-mail: jamunshi@bechtel.com [Bechtel Power Corporation, 5275 Westview Drive, BP2-2C3, Frederick, MD 21703 (United States)

2013-02-15

Highlights: ► A rigorous formulae is proposed to calculate radial stress within prestressed concrete containments. ► The proposed method is validated by finite element analysis in an illustrative practical example. ► A partially prestressed condition is more critical than a fully prestressed condition for radial tension. ► Practical design consideration is provided for detailing of radial reinforcement. -- Abstract: Nuclear containments are critical components for safety of nuclear power plants. Failure can result in catastrophic safety consequences as a result of leakage of radiation. Prestressed concrete containments have been used in large nuclear power plants with significant design internal pressure. These containments are generally reinforced with prestressing tendons in the circumferential (hoop) and meridional (vertical) directions. The curvature effect of the tendons introduces radial tensile stresses in the concrete shell which are generally neglected in the design of such structures. It is assumed that such tensile radial stresses are small as such no radial reinforcement is provided for this purpose. But recent instances of significant delaminations in Crystal River Unit 3 in Florida have elevated the need for reevaluation of the radial tension issue in prestressed containment. Note that currently there are no well accepted industry standards for design and detailing of radial reinforcement. This paper discusses the issue of radial tension in prestressed cylindrical and dome shaped structures and proposes formulae to calculate radial stresses. A practical example is presented to illustrate the use of the proposed method which is then verified by using state of art finite element analysis. This paper also provides some practical design consideration for detailing of radial reinforcement in prestressed containments.

Low enrichment fuel conversion for Iowa State University. Final report

International Nuclear Information System (INIS)

Bullen, D.B.; Wendt, S.E.

1996-01-01

The UTR-10 research and teaching reactor at Iowa State University (ISU) has been converted from high-enriched fuel (HEU) to low- enriched fuel (LEU) under Grant No. DE-FG702-87ER75360 from the Department of Energy (DOE). The original contract period was August 1, 1987 to July 31, 1989. The contract was extended to February 28, 1991 without additional funding. Because of delays in receiving the LEU fuel and the requirement for disassembly of the HEU assemblies, the contract was renewed first through May 31, 1992, then through May 31, 1993 with additional funding, and then again through July 31, 1994 with no additional funding. In mid-August the BMI cask was delivered to Iowa State. Preparations are underway to ship the HEU fuel when NRC license amendments for the cask are approved

Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2008

Energy Technology Data Exchange (ETDEWEB)

Primm, Trent [ORNL; Chandler, David [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Sease, John D [ORNL; Jolly, Brian C [ORNL

2009-03-01

This report documents progress made during FY 2008 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Scoping experiments with various manufacturing methods for forming the LEU alloy profile are presented.

Concept and nuclear performance of direct-enrichment fusion breeder blanket using UO2 powder

International Nuclear Information System (INIS)

Oka, Yoshiaki; Kasahara, Takayasu; An, Shigehiro

1985-01-01

A new concept is presented for direct enrichment of fissile fuel in the blanket of a fusion-fission hybrid reactor. The enriched fuel produced by this means can be used in fission reactors without reprocessing. The outstanding feature of the concept is the powdered form in which UO 2 fuel is placed in the reactor blanket, where it is irradiated to the requisite enrichment for use as fuel in burner reactor, e.g. 3%. After removal from blanket, the powder is mixed to homogenize the enrichment. Fuel pellets and assemblies are then fabricated from the powder without reprocessing. The concept of irradiating UO 2 in powder eliminates the problems of spatial nonuniformity in fissile enrichment, and of radiation damage to fuel clad, encountered in attempting to enrich prefabricated fuel. Powder mixing for homogenization brings the additional benefit of removing volatile fission products. Also burnable poison can be added, as necessary, after irradiation. An extensive neutronic parameter survey showed that the optimum blanket arrangement for this enrichment concept is one presenting a fission suppressing configuration and with beryllium adopted as moderator. By this arrangement, the average 239 Pu enrichment obtained on the natural UO 2 fuel in the blanket reaches 3% after only 0.56 MW.yr/m"2 exposure. A conceptual design is presented of the blanket, together with associated fusion breeder, from which, practical application of the concept is shown to be promising. (author)

Computer model analysis of the radial artery pressure waveform.

Science.gov (United States)

Schwid, H A; Taylor, L A; Smith, N T

1987-10-01

Simultaneous measurements of aortic and radial artery pressures are reviewed, and a model of the cardiovascular system is presented. The model is based on resonant networks for the aorta and axillo-brachial-radial arterial system. The model chosen is a simple one, in order to make interpretation of the observed relationships clear. Despite its simplicity, the model produces realistic aortic and radial artery pressure waveforms. It demonstrates that the resonant properties of the arterial wall significantly alter the pressure waveform as it is propagated from the aorta to the radial artery. Although the mean and end-diastolic radial pressures are usually accurate estimates of the corresponding aortic pressures, the systolic pressure at the radial artery is often much higher than that of the aorta due to overshoot caused by the resonant behavior of the radial artery. The radial artery dicrotic notch is predominantly dependent on the axillo-brachial-radial arterial wall properties, rather than on the aortic valve or peripheral resistance. Hence the use of the radial artery dicrotic notch as an estimate of end systole is unreliable. The rate of systolic upstroke, dP/dt, of the radial artery waveform is a function of many factors, making it difficult to interpret. The radial artery waveform usually provides accurate estimates for mean and diastolic aortic pressures; for all other measurements it is an inadequate substitute for the aortic pressure waveform. In the presence of low forearm peripheral resistance the mean radial artery pressure may significantly underestimate the mean aortic pressure, as explained by a voltage divider model.

Enrichment, isolation and characterization of fungi tolerant to 1-ethyl-3-methylimidazolium acetate

Energy Technology Data Exchange (ETDEWEB)

Singer, S.W.; Reddy, A. P.; Gladden, J. M.; Guo, H.; Hazen, T.C.; Simmons, B. A.; VanderGheynst, J. S.

2010-12-15

This work aims to characterize microbial tolerance to 1-ethyl-3-methylimidazolium acetate ([C2mim][OAc]), ionic liquid that has emerged as a novel biomass pretreatment for lignocellulosic biomass. Enrichment experiments performed using inocula treated with [C2mim][OAc] under solid and liquid cultivation yielded fungal populationsdominated by Aspergilli. Ionic liquid-tolerant Aspergillus isolates from these enrichments were capable of growing in a radial plate growth assay in the presence of 10% [C2mim][OAc]. When a [C2mim][OAc]-tolerant Aspergillus fumigatus strain was grown in the presence of switchgrass, endoglucanases and xylanases were secreted that retained residual enzymatic activity in the presence of 20% [C2mim][OAc]. The results of the study suggest tolerance to ionic liquids is a general property of Aspergilli. Tolerance to an industrially important ionic liquid was discovered in a fungal genera that is widely used in biotechnology, including biomass deconstruction.

Proteomic characterization of golgi membranes enriched from Arabidopsis suspension cell cultures

DEFF Research Database (Denmark)

Hansen, Sara Fasmer; Ebert, Berit; Rautengarten, Carsten

2016-01-01

The plant Golgi apparatus has a central role in the secretory pathway and is the principal site within the cell for the assembly and processing of macromolecules. The stacked membrane structure of the Golgi apparatus along with its interactions with the cytoskeleton and endoplasmic reticulum has...... historically made the isolation and purification of this organelle difficult. Density centrifugation has typically been used to enrich Golgi membranes from plant microsomal preparations, and aside from minor adaptations, the approach is still widely employed. Here we outline the enrichment of Golgi membranes...... from an Arabidopsis cell suspension culture that can be used to investigate the proteome of this organelle. We also provide a useful workflow for the examination of proteomic data as the result of multiple analyses. Finally, we highlight a simple technique to validate the subcellular localization...

Analytical and experimental analysis of YALINA-Booster and YALINA-Thermal assemblies

International Nuclear Information System (INIS)

Kiyavitskaya, H.; Bournos, V.; Mazanik, S.; Khilmanovich, A.; Martsinkevich, B.; Routkovskaya, Ch.; Edchik, I.; Fokov, Y.; Sadovich, S.; Fedorenko, A.; Gohar, Y.; Talamo, A.

2010-01-01

Full text: Accelerator Driven Systems (ADS) may play an important role in future nuclear fuel cycles to reduce the longterm radiotoxicity and volume of spent nuclear fuel. It is proposed that ADS will produce energy and incinerate radioactive waste. This technology was called Accelerator Driven Transmutation Technology (ADTT). The most important problems of this technology are monitoring of a reactivity level in on-line regime, a choice of neutron spectrum appropriate for incineration of Minor Actinides (MA) and transmutation of Long Lived Fission Products (LLFP) and etc. Before the designing and construction of an installation it is necessary to carry out R and D to validate codes, nuclear data libraries and other instrumentations. The YALINA facility is designed to study the ADS physics and to investigate the transmutation reaction rates of MA and LLFP. The main objective of the YALINA benchmark is to compare the results from different calculation methods with each other and experimental data. The benchmark is based on the current YALINA facility configuration, which provides the opportunity to verify the prediction capability of the different methods. The experimental data have been obtained in the frame of the ISTC Projects B1341 'Analytical and experimental evaluation of the possibility to create a universal volume source of neutrons in the sub-critical booster assembly with low enrichment uranium fuel driven by a neutron generator' and B1732P 'Analytical and experimental evaluating the possibility of creation of universal volume source of neutrons in the sub-critical booster assembly with low enriched uranium fuel driven by the neutron generator'. In this paper a comparison of the experimental and calculated data obtained for YALINA-Booster subcritical assembly with a fuel of different enrichment and for YALINA-Thermal with a different number of control rods (216, 245 and 280) will be done.

Fuel-steel mixing and radial mesh effects in power excursion simulations

International Nuclear Information System (INIS)

Chen, X.-N.; Rineiski, A.; Gabrielli, F.; Andriolo, L.; Vezzoni, B.; Li, R.; Maschek, W.; Kiefhaber, E.

2016-01-01

Highlights: • Fuel-steel mixing and radial mesh effects are significant on power excursion. • The earliest power peak is reduced and retarded by these two effects. • Unprotected loss of coolant transients in ESFR core are calculated. - Abstract: This paper deals with SIMMER-III once-through simulations of the earliest power excursion initiated by an unprotected loss of flow (ULOF) in the Working Horse design of the European Sodium Cooled Fast Reactor (ESFR). Since the sodium void effect is strictly positive in this core and dominant in the transient, a power excursion is initiated by sodium boiling in the ULOF case. Two major effects, namely (1) reactivity effects due to fuel-steel mixing after melting and (2) the radial mesh size, which were not considered originally in SIMMER simulations for ESFR, are studied. The first effect concerns the reactivity difference between the heterogeneous fuel/clad/wrapper configuration and the homogeneous mixture of steel and fuel. The full core homogenization (due to melting) effect is −2 $, though a smaller effect takes place in case of partial core melting. The second effect is due to the SIMMER sub-assembly (SA) coarse mesh treatment, where a simultaneous sodium boiling onset in all SAs belonging to one ring leads to an overestimated reactivity ramp. For investigating the influence of fuel/steel mixing effects, a lumped “homogenization” reactivity feedback has been introduced, being proportional to the molten steel mass. For improving the coarse mesh treatment, we employ finer radial meshes to take the subchannel effects into account, where the side and interior channels have different coolant velocities and temperatures. The simulation results show that these two effects have significant impacts on the earliest power excursion after the sodium boiling.

The Bare Critical Assembly of Natural Uranium and Heavy Water

Energy Technology Data Exchange (ETDEWEB)

Popovic, D [Boris Kidric Institute of Nuclear Sciences, Vinca, Belgrade (Yugoslavia)

1958-07-01

The first reactor built in Yugoslavia was the bare zero energy heavy water and natural uranium assembly at the Boris Kidric Institute of Nuclear Sciences, Belgrade. The reactor went critical on April 29, 1958. The possession of four tons of natural uranium metal and the temporary availability of seven tons of heavy water encouraged the staff of the Institute to build a critical assembly. A critical assembly was chosen, rather than high flux reactor, because the heavy water was available only temporarily. Besides, a 10 MW, enriched uranium, research reactor is being built at the same Institute and should be ready for operation late this year. It was supposed that the zero energy reactor would provide experience in carrying out critical experiments, operational experience with nuclear reactors, and the possibility for an extensive program in reactor physics. (author)

A Mixed-Oxide Assembly Design for Rapid Disposition of Weapons Plutonium in Pressurized Water Reactors

International Nuclear Information System (INIS)

Alonso, Gustavo; Adams, Marvin L.

2002-01-01

We have created a new mixed-oxide (MOX) fuel assembly design for standard pressurized water reactors (PWRs). Design goals were to maximize the plutonium throughput while introducing the lowest perturbation possible to the control and safety systems of the reactor. Our assembly design, which we call MIX-33, offers some advantages for the disposition of weapons-grade plutonium; it increases the disposition rate by 8% while increasing the worth of control material, compared to a previous Westinghouse design. The MIX-33 design is based upon two ideas: the use of both uranium and plutonium fuel pins in the same assembly, and the addition of water holes in the assembly. The main result of this paper is that both of these ideas are effective at increasing Pu throughput and increasing the worth of control material. With this new design, according to our analyses, we can transition smoothly from a full low-enriched-uranium (LEU) core to a full MIX-33 core while meeting the operational and safety requirements of a standard PWR. Given an interruption of the MOX supply, we can transition smoothly back to full LEU while meeting safety margins and using standard LEU assemblies with uniform pinwise enrichment distribution. If the MOX supply is interrupted for only one cycle, the transition back to a full MIX-33 core is not as smooth; high peaking could cause power to be derated by a few percent for a few weeks at the beginning of one transition cycle

MR accuracy and arthroscopic incidence of meniscal radial tears

Energy Technology Data Exchange (ETDEWEB)

Magee, Thomas; Shapiro, Marc; Williams, David [Department of Radiology, Neuroimaging Institute, 27 East Hibiscus Blvd., Melbourne, FL 32901 (United States)

2002-12-01

A meniscal radial tear is a vertical tear that involves the inner meniscal margin. The tear is most frequent in the middle third of the lateral meniscus and may extend outward in any direction. We report (1) the arthroscopic incidence of radial tears, (2) MR signs that aid in the detection of radial tears and (3) our prospective accuracy in detection of radial tears. Design and patients. Three musculoskeletal radiologists prospectively read 200 consecutive MR examinations of the knee that went on to arthroscopy by one orthopedic surgeon. MR images were assessed for location and MR characteristics of radial tears. MR criteria used for diagnosis of a radial tear were those outlined by Tuckman et al.: truncation, abnormal morphology and/or lack of continuity or absence of the meniscus on one or more MR images. An additional criterion used was abnormal increased signal in that area on fat-saturated proton density or T2-weighted coronal and sagittal images. Prospective MR readings were correlated with the arthroscopic findings.Results. Of the 200 consecutive knee arthroscopies, 28 patients had radial tears reported arthroscopically (14% incidence). MR readings prospectively demonstrated 19 of the 28 radial tears (68% sensitivity) when the criteria for diagnosis of a radial tear were truncation or abnormal morphology of the meniscus. With the use of the additional criterion of increased signal in the area of abnormal morphology on fat-saturated T2-weighted or proton density weighted sequences, the prospective sensitivity was 25 of 28 radial tears (89% sensitivity). There were no radial tears described in MR reports that were not demonstrated on arthroscopy (i.e., there were no false positive MR readings of radial tears in these 200 patients). Radial tears are commonly seen at arthroscopy. There was a 14% incidence in this series of 200 patients who underwent arthroscopy. Prospective detection of radial tears was 68% as compared with arthroscopy when the criteria as

MR accuracy and arthroscopic incidence of meniscal radial tears

International Nuclear Information System (INIS)

Magee, Thomas; Shapiro, Marc; Williams, David

2002-01-01

A meniscal radial tear is a vertical tear that involves the inner meniscal margin. The tear is most frequent in the middle third of the lateral meniscus and may extend outward in any direction. We report (1) the arthroscopic incidence of radial tears, (2) MR signs that aid in the detection of radial tears and (3) our prospective accuracy in detection of radial tears. Design and patients. Three musculoskeletal radiologists prospectively read 200 consecutive MR examinations of the knee that went on to arthroscopy by one orthopedic surgeon. MR images were assessed for location and MR characteristics of radial tears. MR criteria used for diagnosis of a radial tear were those outlined by Tuckman et al.: truncation, abnormal morphology and/or lack of continuity or absence of the meniscus on one or more MR images. An additional criterion used was abnormal increased signal in that area on fat-saturated proton density or T2-weighted coronal and sagittal images. Prospective MR readings were correlated with the arthroscopic findings.Results. Of the 200 consecutive knee arthroscopies, 28 patients had radial tears reported arthroscopically (14% incidence). MR readings prospectively demonstrated 19 of the 28 radial tears (68% sensitivity) when the criteria for diagnosis of a radial tear were truncation or abnormal morphology of the meniscus. With the use of the additional criterion of increased signal in the area of abnormal morphology on fat-saturated T2-weighted or proton density weighted sequences, the prospective sensitivity was 25 of 28 radial tears (89% sensitivity). There were no radial tears described in MR reports that were not demonstrated on arthroscopy (i.e., there were no false positive MR readings of radial tears in these 200 patients). Radial tears are commonly seen at arthroscopy. There was a 14% incidence in this series of 200 patients who underwent arthroscopy. Prospective detection of radial tears was 68% as compared with arthroscopy when the criteria as

Reaction Rate Distributions and Ratios in FR0 Assemblies 1, 2 and 3

Energy Technology Data Exchange (ETDEWEB)

Andersson, T L

1966-06-15

The spatial distribution of different reaction rates and reaction ratios in Assemblies 1, 2 and 3 of the fast reactor FR0 was measured by fission chamber scans and foil activation technique. Assemblies 1 and 2 had cores of undiluted fuel (uranium metal enriched to 20 % U{sup 235}) while the core of Assembly 3 was diluted with about 30 vol. % graphite. All the systems had a thick copper reflector, The experimental results were compared with calculated values obtained from DSN and TDC multigroup spectra and group cross-section sets for the reactions. Good agreement between experiment and calculations is generally obtained in the core region but in the reflector the neutron spectrum is calculated too hard.

Reorganization of a dense granular assembly: The unjamming response function

Science.gov (United States)

Kolb, Évelyne; Cviklinski, Jean; Lanuza, José; Claudin, Philippe; Clément, Éric

2004-03-01

We investigate the mechanical properties of a static dense granular assembly in response to a local forcing. To this end, a small cyclic displacement is applied on a grain in the bulk of a two-dimensional disordered packing under gravity and the displacement fields are monitored. We evidence a dominant long range radial response in the upper half part above the solicitation and after a large number of cycles the response is “quasireversible” with a remanent dissipation field exhibiting long range streams and vortexlike symmetry.

Radial velocities of RR Lyrae stars

International Nuclear Information System (INIS)

Hawley, S.L.; Barnes, T.G. III

1985-01-01

283 spectra of 57 RR Lyrae stars have been obtained using the 2.1-m telescope at McDonald Observatory. Radial velocities were determined using a software cross-correlation technique. New mean radial velocities were determined for 46 of the stars. 11 references

Search Greedy for radial fuel optimization

International Nuclear Information System (INIS)

Ortiz, J. J.; Castillo, J. A.; Pelta, D. A.

2008-01-01

In this work a search algorithm Greedy is presented for the optimization of fuel cells in reactors BWR. As first phase a study was made of sensibility of the Factor of Pick of Local Power (FPPL) of the cell, in function of the exchange of the content of two fuel rods. His way it could settle down that then the rods to exchange do not contain gadolinium, small changes take place in the value of the FPPL of the cell. This knowledge was applied later in the search Greedy to optimize fuel cell. Exchanges of rods with gadolinium takes as a mechanism of global search and exchanges of rods without gadolinium takes as a method of local search. It worked with a cell of 10x10 rods and 2 circular water channels in center of the same one. From an inventory of enrichments of uranium and concentrations of given gadolinium and one distribution of well-known enrichments; the techniques finds good solutions that the FPPL minimizes, maintaining the factor of multiplication of neutrons in a range appropriate of values. In the low part of the assembly of a lot of recharge of a cycle of 18 months the cells were placed. The values of FPPL of the opposing cells are similar or smaller to those of the original cell and with behaviors in the nucleus also comparable to those obtained with the original cell. The evaluation of the cells was made with the code of transport CASMO-IV and the evaluation of the nucleus was made by means of the one simulator of the nucleus SIMULATE-3. (Author)

Multiple Surrogate Modeling for Wire-Wrapped Fuel Assembly Optimization

International Nuclear Information System (INIS)

Raza, Wasim; Kim, Kwang-Yong

2007-01-01

In this work, shape optimization of seven pin wire wrapped fuel assembly has been carried out in conjunction with RANS analysis in order to evaluate the performances of surrogate models. Previously, Ahmad and Kim performed the flow and heat transfer analysis based on the three-dimensional RANS analysis. But numerical optimization has not been applied to the design of wire-wrapped fuel assembly, yet. Surrogate models are being widely used in multidisciplinary optimization. Queipo et al. reviewed various surrogates based models used in aerospace applications. Goel et al. developed weighted average surrogate model based on response surface approximation (RSA), radial basis neural network (RBNN) and Krigging (KRG) models. In addition to the three basic models, RSA, RBNN and KRG, the multiple surrogate model, PBA also has been employed. Two geometric design variables and a multi-objective function with a weighting factor have been considered for this problem

Subcritical nuclear assembly

Energy Technology Data Exchange (ETDEWEB)

Vega C, H. R., E-mail: fermineutron@yahoo.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico)

2014-08-15

A Subcritical Nuclear Assembly is a device where the nuclear-fission chain reaction is initiated and maintained using an external neutron source. It is a valuable educational and research tool where in a safe way many reactor parameters can be measured. Here, we have used the Wigner-Seitz method in the six-factor formula to calculate the effective multiplication factor of a subcritical nuclear reactor Nuclear Chicago model 9000. This reactor has approximately 2500 kg of natural uranium heterogeneously distributed in slugs. The reactor uses a {sup 239}PuBe neutron source that is located in the center of an hexagonal array. Using Monte Carlo methods, with the MCNP5 code, a three-dimensional model of the subcritical reactor was designed to estimate the effective multiplication factor, the neutron spectra, the total and thermal neutron fluences along the radial and axial axis. With the neutron spectra in two locations outside the reactor the ambient dose equivalent were estimated. (Author)

Subcritical nuclear assembly

International Nuclear Information System (INIS)

Vega C, H. R.

2014-08-01

A Subcritical Nuclear Assembly is a device where the nuclear-fission chain reaction is initiated and maintained using an external neutron source. It is a valuable educational and research tool where in a safe way many reactor parameters can be measured. Here, we have used the Wigner-Seitz method in the six-factor formula to calculate the effective multiplication factor of a subcritical nuclear reactor Nuclear Chicago model 9000. This reactor has approximately 2500 kg of natural uranium heterogeneously distributed in slugs. The reactor uses a 239 PuBe neutron source that is located in the center of an hexagonal array. Using Monte Carlo methods, with the MCNP5 code, a three-dimensional model of the subcritical reactor was designed to estimate the effective multiplication factor, the neutron spectra, the total and thermal neutron fluences along the radial and axial axis. With the neutron spectra in two locations outside the reactor the ambient dose equivalent were estimated. (Author)

High-performance 16-way Ku-band radial power combiner based on the TE01-circular waveguide mode

Science.gov (United States)

Montejo-Garai, José R.; Saracho-Pantoja, Irene O.; Ruiz-Cruz, Jorge A.; Rebollar, Jesús M.

2018-03-01

This work presents a 16-way Ku-band radial power combiner for high power and high frequency applications, using the very low loss TE01 circular waveguide mode. The accomplished design shows an excellent performance: the experimental prototype has a return loss better than 30 dB, with a balance for the amplitudes of (±0.15 dB) and (±2.5°) for the phases, in a 16.7% fractional bandwidth (2 GHz centered at 12 GHz). For obtaining these outstanding specifications, required, for instance, in high-frequency amplification or on plasma systems, a rigorous step-by-step procedure is presented. First, a high-purity mode transducer has been designed, from the TE10 mode in the rectangular waveguide to the TE01 mode in the circular waveguide, with very high attenuation (>50 dB) for the other propagating and evanescent modes in the circular waveguide. This transducer has been manufactured and measured in a back-to-back configuration, validating the design process. Second, an E-plane 16-way radial power divider has been designed, where the power is coupled from the 16 non-reduced-height radial standard waveguides into the TE01 circular waveguide mode, improving the insertion loss response and removing the usual tapered transformers of previous designs limiting the power handling. Finally, both the transducer and the divider have been assembled to make the final radial combiner. The prototype has been carefully manufactured, showing very good agreement between the measurements and the full-wave simulations.

The slightly-enriched spectral shift control reactor

Energy Technology Data Exchange (ETDEWEB)

Martin, W.R.; Lee, J.C.; Larsen, E.W. (Michigan Univ., Ann Arbor, MI (United States). Dept. of Nuclear Engineering); Edlund, M.C. (Virginia Polytechnic Inst. and State Univ., Blacksburg, VA (United States). Dept. of Mechanical and Nuclear Engineering)

1991-11-01

An advanced converter reactor design utilizing mechanical spectral shift control rods in a conventional pressurized water reactor configuration is under investigation. The design is based on the principle that a harder spectrum during the early part of the fuel cycle will result in large neutron captures in fertile {sup 238}U, which can then be burned in situ in a softer spectrum later in the cycle. Preliminary design calculations performed during FY 89 showed that the slightly-enriched spectral shift reactor design offers the benefit of substantially increased fuel resource utilization with the proven safety characteristics of the pressurized water reactor technology retained. Optimization of the fuel design and development of fuel management strategies were carried out in FY 90, along with effort to develop and validate neutronic methodology for tight-lattice configurations with hard spectra. During FY 91, the final year of the grant, the final Slightly-Enriched Spectral Shift Reactor (SESSR) design was determined, and reference design analyses were performed for the assemblies as well as the global core configuration, both at the beginning of cycle (BOC) and with depletion. The final SESSR design results in approximately a 20% increase in the utilization of uranium resources, based on equilibrium fuel cycle analyses. Acceptable pin power peaking is obtained with the final core design, with assembly peaking factors equal to less than 1.04 for spectral shift control rods both inserted and withdrawn, and global peaking factors at BOC predicted to be 1.4. In addition, a negative Moderation Temperature Coefficient (MTC) is maintained for BOC, which is difficult to achieve with conventional advanced converter designs based on a closed fuel cycle. The SESSR design avoids the need for burnable poison absorber, although they could be added if desired to increase the cycle length while maintaining a negative MTC.

The slightly-enriched spectral shift control reactor

International Nuclear Information System (INIS)

Martin, W.R.; Lee, J.C.; Larsen, E.W.; Edlund, M.C.

1991-11-01

An advanced converter reactor design utilizing mechanical spectral shift control rods in a conventional pressurized water reactor configuration is under investigation. The design is based on the principle that a harder spectrum during the early part of the fuel cycle will result in large neutron captures in fertile 238 U, which can then be burned in situ in a softer spectrum later in the cycle. Preliminary design calculations performed during FY 89 showed that the slightly-enriched spectral shift reactor design offers the benefit of substantially increased fuel resource utilization with the proven safety characteristics of the pressurized water reactor technology retained. Optimization of the fuel design and development of fuel management strategies were carried out in FY 90, along with effort to develop and validate neutronic methodology for tight-lattice configurations with hard spectra. During FY 91, the final year of the grant, the final Slightly-Enriched Spectral Shift Reactor (SESSR) design was determined, and reference design analyses were performed for the assemblies as well as the global core configuration, both at the beginning of cycle (BOC) and with depletion. The final SESSR design results in approximately a 20% increase in the utilization of uranium resources, based on equilibrium fuel cycle analyses. Acceptable pin power peaking is obtained with the final core design, with assembly peaking factors equal to less than 1.04 for spectral shift control rods both inserted and withdrawn, and global peaking factors at BOC predicted to be 1.4. In addition, a negative Moderation Temperature Coefficient (MTC) is maintained for BOC, which is difficult to achieve with conventional advanced converter designs based on a closed fuel cycle. The SESSR design avoids the need for burnable poison absorber, although they could be added if desired to increase the cycle length while maintaining a negative MTC

Radial MR images of the knee

International Nuclear Information System (INIS)

Hewes, R.C.; Miller, T.R.

1988-01-01

To profile optimally each portion of the meniscus, the authors use the multiangle, multisection feature of a General Electric SIGNA 1.5-T imager to produce radial images centered on each meniscus. A total of 12-15 sections are imaged at 10 0 -15 0 intervals of each meniscus, yielding perpendicular images of the entire meniscus, comparable with the arthrographic tangential views. The authors review their technique and demonstrate correlation cases between the radial gradient recalled acquisition in a steady state sequences, sagittal and coronal MR images, and arthrograms. Radial images should be a routine part of knee MR imaging

Improvement of operation efficiency for WWER-440 and WWER-1000 for TRIGON fuel assembly design features

Energy Technology Data Exchange (ETDEWEB)

Silberstein, A [European WWER Fuels GmbH, Lyon (France)

1994-12-31

TRIGON 440 and TRIGON 1000 fuel assemblies and their assembly matching counterparts are described. Their role in increasing the efficiency of WWER reactors is stressed. Special attention is paid to their design features as well as calibrated means of predicting behaviour under irradiation from light water reactor core operation. They reduce the fuel cycle cost as a result of the reduced need for natural uranium which have to be enriched and of the smaller number of fuel assemblies which have to be fabricated, stored or reprocessed. The improved control assemblies bring comfort to the plant operator due to intrinsic progress in safety with respect to accidental situation, trouble-free behaviour and long time utilization in the reactor. 14 figs.

21 CFR 866.4800 - Radial immunodiffusion plate.

Science.gov (United States)

2010-04-01

...) MEDICAL DEVICES IMMUNOLOGY AND MICROBIOLOGY DEVICES Immunology Laboratory Equipment and Reagents § 866.4800 Radial immunodiffusion plate. (a) Identification. A radial immunodiffusion plate for clinical use...

Study of low leakage reload schedulle without burnable posion for Angra-1

International Nuclear Information System (INIS)

Sakai, M.; Dias, A.

1989-01-01

At the moment, there is a world trend to design larger cycles for PWR. Then the reload batches are increased, the enrichment in 235 U is increased and/or advanced fuel management strategies with radial low neutron leakage are applied. For the low leakage reloads of Angra-1 calculations were performed for different number of fuel assemblies for reaload batch, 32,36,40,44 and 48, from the 4th cycle up to equilibrium cycle for two different enrichments 3,4 W/O and 3,9 W/O in 235 U. The results showed that for the enrichments used without burnable posion it is possible to reach an increase in cycle lenghts between 3% and 8% for the same conditions. (author) [pt

Stirling Engine With Radial Flow Heat Exchangers

Science.gov (United States)

Vitale, N.; Yarr, George

1993-01-01

Conflict between thermodynamical and structural requirements resolved. In Stirling engine of new cylindrical configuration, regenerator and acceptor and rejector heat exchangers channel flow of working gas in radial direction. Isotherms in regenerator ideally concentric cylinders, and gradient of temperature across regenerator radial rather than axial. Acceptor and rejector heat exchangers located radially inward and outward of regenerator, respectively. Enables substantial increase in power of engine without corresponding increase in diameter of pressure vessel.

Diverted assembly for radioactive material

International Nuclear Information System (INIS)

Andrews, K.M.; Starenchak, R.W.

1989-01-01

This patent describes a diverter assembly for diverting a pneumatically conveyed holder for radioactive material between a central conveying tube and one of a plurality of conveying tubes radially offset from the central conveying tube. It comprises: an airtight container having a hollow interior, a first aperture about which the central tube is connected, and a respective plurality of second apertures about which respective offset tubes are connected; a diverter tube in the container having a first end located immediately adjacent the first aperture and a second end offset from the first end a distance equal to the radial offset of the conveying tubes from the central tube; a first mounting means in the container for mounting the diverter tube for rotation about a longitudinal axis of the first end such that the second end is selectively brought into alignment with respective the second apertures; a rotary seal means for sealing the first end of the diverter tube from the interior of the container during and after rotation of the diverter tube; a spring biased seal means for sealing the second end of the diverter tube from the interior of the container during and after rotation of the diverter tube to a selected second aperture; an indexing means for rotatable indexing the second end of the diverter tube; and a drive means for selectively driving the indexing means

Uranium enrichment

International Nuclear Information System (INIS)

1990-01-01

This report looks at the following issues: How much Soviet uranium ore and enriched uranium are imported into the United States and what is the extent to which utilities flag swap to disguise these purchases? What are the U.S.S.R.'s enriched uranium trading practices? To what extent are utilities required to return used fuel to the Soviet Union as part of the enriched uranium sales agreement? Why have U.S. utilities ended their contracts to buy enrichment services from DOE?

Carbon nanofibers with radially grown graphene sheets derived from electrospinning for aqueous supercapacitors with high working voltage and energy density

Science.gov (United States)

Zhao, Lei; Qiu, Yejun; Yu, Jie; Deng, Xianyu; Dai, Chenglong; Bai, Xuedong

2013-05-01

Improvement of energy density is an urgent task for developing advanced supercapacitors. In this paper, aqueous supercapacitors with high voltage of 1.8 V and energy density of 29.1 W h kg-1 were fabricated based on carbon nanofibers (CNFs) and Na2SO4 electrolyte. The CNFs with radially grown graphene sheets (GSs) and small average diameter down to 11 nm were prepared by electrospinning and carbonization in NH3. The radially grown GSs contain between 1 and a few atomic layers with their edges exposed on the surface. The CNFs are doped with nitrogen and oxygen with different concentrations depending on the carbonizing temperature. The supercapacitors exhibit excellent cycling performance with the capacity retention over 93.7% after 5000 charging-discharging cycles. The unique structure, possessing radially grown GSs, small diameter, and heteroatom doping of the CNFs, and application of neutral electrolyte account for the high voltage and energy density of the present supercapacitors. The present supercapacitors are of high promise for practical application due to the high energy density and the advantages of neutral electrolyte including low cost, safety, low corrosivity, and convenient assembly in air.

Beyond assembly bias: exploring secondary halo biases for cluster-size haloes

Science.gov (United States)

Mao, Yao-Yuan; Zentner, Andrew R.; Wechsler, Risa H.

2018-03-01

Secondary halo bias, commonly known as `assembly bias', is the dependence of halo clustering on a halo property other than mass. This prediction of the Λ Cold Dark Matter cosmology is essential to modelling the galaxy distribution to high precision and interpreting clustering measurements. As the name suggests, different manifestations of secondary halo bias have been thought to originate from halo assembly histories. We show conclusively that this is incorrect for cluster-size haloes. We present an up-to-date summary of secondary halo biases of high-mass haloes due to various halo properties including concentration, spin, several proxies of assembly history, and subhalo properties. While concentration, spin, and the abundance and radial distribution of subhaloes exhibit significant secondary biases, properties that directly quantify halo assembly history do not. In fact, the entire assembly histories of haloes in pairs are nearly identical to those of isolated haloes. In general, a global correlation between two halo properties does not predict whether or not these two properties exhibit similar secondary biases. For example, assembly history and concentration (or subhalo abundance) are correlated for both paired and isolated haloes, but follow slightly different conditional distributions in these two cases. This results in a secondary halo bias due to concentration (or subhalo abundance), despite the lack of assembly bias in the strict sense for cluster-size haloes. Due to this complexity, caution must be exercised in using any one halo property as a proxy to study the secondary bias due to another property.

Regulation of corneal stroma extracellular matrix assembly.

Science.gov (United States)

Chen, Shoujun; Mienaltowski, Michael J; Birk, David E

2015-04-01

The transparent cornea is the major refractive element of the eye. A finely controlled assembly of the stromal extracellular matrix is critical to corneal function, as well as in establishing the appropriate mechanical stability required to maintain corneal shape and curvature. In the stroma, homogeneous, small diameter collagen fibrils, regularly packed with a highly ordered hierarchical organization, are essential for function. This review focuses on corneal stroma assembly and the regulation of collagen fibrillogenesis. Corneal collagen fibrillogenesis involves multiple molecules interacting in sequential steps, as well as interactions between keratocytes and stroma matrix components. The stroma has the highest collagen V:I ratio in the body. Collagen V regulates the nucleation of protofibril assembly, thus controlling the number of fibrils and assembly of smaller diameter fibrils in the stroma. The corneal stroma is also enriched in small leucine-rich proteoglycans (SLRPs) that cooperate in a temporal and spatial manner to regulate linear and lateral collagen fibril growth. In addition, the fibril-associated collagens (FACITs) such as collagen XII and collagen XIV have roles in the regulation of fibril packing and inter-lamellar interactions. A communicating keratocyte network contributes to the overall and long-range regulation of stromal extracellular matrix assembly, by creating micro-domains where the sequential steps in stromal matrix assembly are controlled. Keratocytes control the synthesis of extracellular matrix components, which interact with the keratocytes dynamically to coordinate the regulatory steps into a cohesive process. Mutations or deficiencies in stromal regulatory molecules result in altered interactions and deficiencies in both transparency and refraction, leading to corneal stroma pathobiology such as stromal dystrophies, cornea plana and keratoconus. Copyright © 2014 Elsevier Ltd. All rights reserved.

Neutronics substantiation of possibility for conversion of the WWR-K reactor core to operation with low-enriched fuel

International Nuclear Information System (INIS)

Arinkin, F.M.; Gizatulin, Sh.H.; Zhantikin, T.M.; Koltochnik, S.N.; Takibaev, A.Zh.; Talanov, S.V.; Chakrov, P.V.; Chekushina, L.V.

2002-01-01

The studies are aimed to calculation and experimental justification of possibility for conversion of the WWR-R reactor core to low-enriched nuclear fuel (the 19.75-% enrichment in isotope U-235), resulting in reducing the risk of non-sanctioned proliferation of nuclear materials which can be used as weapons materials. The analysis of available published data, related to problem of reduction of enrichment in the fuel used in research thermal reactors, has been carried out. Basing on the analysis results, reference fuel compositions have been chosen, in particular, uranium dioxide (UO 2 ) in aluminum master form and the UA1 4 alloy. Preliminary calculations have shown that, with the WWR-K reactor core preserved existing critical characteristics (the fuel composition: UA1 4 ), the uranium concentration in the fuel element is to be increased by a factor of 2.0-2.2, being impossible technologically. The calculations have been performed by means of the Monte Carlo computational codes. The program of optimal conversion of the WWR-K reactor core to low-enriched fuel has been developed, including: development of calculation models of the reactor core, composed of various designs of fuel elements and fuel assemblies (FA), on a base of corresponding computational codes (diffusion, statistical, etc.); implementation of experiments in the zero-power reactor (critical assembly) with the WWR-C-type FA, in view of correction of the computational constants used in calculations; implementation of reactor core neutronics calculations, in view of selection of the U-235 optimal content in the low-enriched fuel elements and choice of FA reload strategy at the regime of reactor core after burning; determination of the fuel element specification; determination of the critical and operational loads for the reactor core composed of rod/tubular fuel elements; calculation of the efficiency of the protection control system effectors, optimization of its composition, number and locations in the

Radial pattern of nuclear decay processes

International Nuclear Information System (INIS)

Iskra, W.; Mueller, M.; Rotter, I.; Technische Univ. Dresden

1994-05-01

At high level density of nuclear states, a separation of different time scales is observed (trapping effect). We calculate the radial profile of partial widths in the framework of the continuum shell model for some 1 - resonances with 2p-2h nuclear structure in 16 O as a function of the coupling strength to the continuum. A correlation between the lifetime of a nuclear state and the radial profile of the corresponding decay process is observed. We conclude from our numerical results that the trapping effect creates structures in space and time characterized by a small radial extension and a short lifetime. (orig.)

Self-assembled magnetic filter for highly efficient immunomagnetic separation.

Science.gov (United States)

Issadore, David; Shao, Huilin; Chung, Jaehoon; Newton, Andita; Pittet, Mikael; Weissleder, Ralph; Lee, Hakho

2011-01-07

We have developed a compact and inexpensive microfluidic chip, the self-assembled magnetic filter, to efficiently remove magnetically tagged cells from suspension. The self-assembled magnetic filter consists of a microfluidic channel built directly above a self-assembled NdFeB magnet. Micrometre-sized grains of NdFeB assemble to form alternating magnetic dipoles, creating a magnetic field with a very strong magnitude B (from the material) and field gradient ▽B (from the configuration) in the microfluidic channel. The magnetic force imparted on magnetic beads is measured to be comparable to state-of-the-art microfabricated magnets, allowing for efficient separations to be performed in a compact, simple device. The efficiency of the magnetic filter is characterized by sorting non-magnetic (polystyrene) beads from magnetic beads (iron oxide). The filter enriches the population of non-magnetic beads to magnetic beads by a factor of >10(5) with a recovery rate of 90% at 1 mL h(-1). The utility of the magnetic filter is demonstrated with a microfluidic device that sorts tumor cells from leukocytes using negative immunomagnetic selection, and concentrates the tumor cells on an integrated membrane filter for optical detection.

Gas-phase UF6 enrichment monitor for enrichment plant safeguards

International Nuclear Information System (INIS)

Strittmatter, R.B.; Tape, J.W.

1980-03-01

An in-line enrichment monitor is being developed to provide real-time enrichment data for the gas-phase UF 6 feed stream of an enrichment plant. The nondestructive gamma-ray assay method can be used to determine the enrichment of natural UF 6 with a relative precision of better than 1% for a wide range of pressures

Intraluminal milrinone for dilation of the radial artery graft.

Science.gov (United States)

García-Rinaldi, R; Soltero, E R; Carballido, J; Mojica, J

1999-01-01

There is renewed interest in the use of the radial artery as a conduit for coronary artery bypass surgery. The radial artery is, however, a very muscular artery, prone to vasospasm. Milrinone, a potent vasodilator, has demonstrated vasodilatory properties superior to those of papaverine. In this report, we describe our technique of radial artery harvesting and the adjunctive use of intraluminal milrinone as a vasodilator in the preparation of this conduit for coronary artery bypass grafting. We have used these techniques in 25 patients who have undergone coronary artery bypass grafting using the radial artery. No hand ischemic complications have been observed in this group. Intraluminal milrinone appears to dilate and relax the radial artery, rendering this large conduit spasm free and very easy to use. We recommend the skeletonization technique for radial artery harvesting and the use of intraluminal milrinone as a radial artery vasodilator in routine myocardial revascularization. PMID:10524740

Multi-recycling of transuranic elements in a PWR assembly with reduced fuel rod diameter

International Nuclear Information System (INIS)

Chambers, Alex; Ragusa, Jean C.

2014-01-01

Highlights: • Study of multiple recycling passes of transuranic elements: (a) without exceeding 5 wt.% on U-235 enrichment; (b) using PWR fuel assemblies compatible with current reactor core internals. • Isotopic concentrations tend towards an equilibrium after 15 recycle passes, suggesting that thermal recycling may be continued beyond that point. • Radiotoxicity comparisons for once-through UOX, once-recycle MOX-Pu, and multiple recycle passes of MOX-PuNpAm and MOX-PuNpAmCm are presented. - Abstract: This paper examines the multi-recycling of transuranic (TRU) elements (Pu-Np-Am-Cm) in standard Pressurized Water Reactor (PWR) assemblies. The original feed of TRU comes from legacy spent UOX fuel. For all subsequent recycling passes, TRU elements from the previous generation are employed, supplemented by TRU from legacy UOX fuel, as needed. The design criteria include: 235 U enrichment requirements to remain below 5 w/o, TRU loading limits to avoid return to criticality under voided conditions, and assembly power peaking factors. In order to carry out multiple recycling passes within the design envelope, additional neutron moderation is required and achieved by reducing the fuel pellet diameter by about 13%, thus keeping the assembly design compatible with current PWR core internals. TRU transmutation rates and long-term ingestion radiotoxicity results are presented for 15 recycling passes and compared to standard UOX and MOX once-through cycles. The results also show that TRU fuel isotopics and radiotoxicity tend towards an equilibrium, enabling further additional recycling passes

Positive responses of belowground C dynamics to nitrogen enrichment in China.

Science.gov (United States)

Deng, Lei; Peng, Changhui; Zhu, Guangyu; Chen, Lei; Liu, Yulin; Shangguan, Zhouping

2018-03-01

Determining how nitrogen (N) impacts ecosystem carbon (C) cycling is critical to using C sequestration to offset anthropogenic CO 2 emissions. The N deposition rate in China is higher than the global average; however, many results of N enrichment experiments in China have not been included in global syntheses. In this study, we assembled a large dataset that comprised 124 published studies concerning N addition experiments, including 570 observations at 127 sites across China, to quantify the responses of belowground C dynamics to N enrichment in terrestrial ecosystems in China by a meta-analysis. The results showed that overall soil organic C, dissolved organic C (DOC) and soil microbial biomass C (MBC) increased by 1.8, 7.4, and 8.8%, respectively (Penrichment; belowground biomass and litter increased by 14.6 and 24.4%, respectively (Penrichment promoted C inputs into the soil mainly by increasing litter and belowground biomass inputs. Additionally, N enrichment increased C output by increasing soil respiration. Land use type and N addition level had different impacts on the soil C pool and on soil respiration. DOC, MBC, and litter exhibited more positive responses to N deposition in cooler and more arid regions than in other regions. The meta-analysis indicated that N enrichment had a positive impact on belowground C cycles in China. Climate played a greater role than did N deposition level in affecting processes of ecosystem C cycling. Moreover, belowground C cycle processes are determined by complicated interactions among land use type, N enrichment, and climate. Copyright © 2017 Elsevier B.V. All rights reserved.

Channeling of protons through radial deformed carbon nanotubes

Energy Technology Data Exchange (ETDEWEB)

Borka Jovanović, V., E-mail: vborka@vinca.rs [Atomic Physics Laboratory (040), Vinča Institute of Nuclear Sciences, University of Belgrade, P.O. Box 522, 11001 Belgrade (Serbia); Borka, D. [Atomic Physics Laboratory (040), Vinča Institute of Nuclear Sciences, University of Belgrade, P.O. Box 522, 11001 Belgrade (Serbia); Galijaš, S.M.D. [Faculty of Physics, University of Belgrade, P.O. Box 368, 11001 Belgrade (Serbia)

2017-05-18

Highlights: • For the first time we presented theoretically obtained distributions of channeled protons with radially deformed SWNT. • Our findings indicate that influence of the radial deformation is very strong and it should not be omitted in simulations. • We show that the spatial and angular distributions depend strongly of level of radial deformation of nanotube. • Our obtained results can be compared with measured distributions to reveal the presence of various types of defects in SWNT. - Abstract: In this paper we have presented a theoretical investigation of the channeling of 1 GeV protons with the radial deformed (10, 0) single-wall carbon nanotubes (SWNTs). We have calculated channeling potential within the deformed nanotubes. For the first time we presented theoretically obtained spatial and angular distributions of channeled protons with radially deformed SWNT. We used a Monte Carlo (MC) simulation technique. We show that the spatial and angular distributions depend strongly of level of radial deformation of nanotube. These results may be useful for nanotube characterization and production and guiding of nanosized ion beams.

Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

Energy Technology Data Exchange (ETDEWEB)

Renfro, David [ORNL; Chandler, David [ORNL; Cook, David [ORNL; Ilas, Germina [ORNL; Jain, Prashant [ORNL; Valentine, Jennifer [ORNL

2014-10-30

Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy’s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully converted using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the “complex” aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The

Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

Energy Technology Data Exchange (ETDEWEB)

Renfro, David G [ORNL; Chandler, David [ORNL; Cook, David Howard [ORNL; Ilas, Germina [ORNL; Jain, Prashant K [ORNL; Valentine, Jennifer R [ORNL

2014-11-01

Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully converted using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the complex aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The present

Numerical investigation of a heat transfer within the prismatic fuel assembly of a very high temperature reactor

International Nuclear Information System (INIS)

Tak, Nam-il; Kim, Min-Hwan; Lee, Won Jae

2008-01-01

The complex geometry of the hexagonal fuel blocks of the prismatic fuel assembly in a very high temperature reactor (VHTR) hinders accurate evaluations of the temperature profile within the fuel assembly without elaborate numerical calculations. Therefore, simplified models such as a unit cell model have been widely applied for the analyses and designs of prismatic VHTRs since they have been considered as effective approaches reducing the computational efforts. In a prismatic VHTR, however, the simplified models cannot consider a heat transfer within a fuel assembly as well as a coolant flow through a bypass gap between the fuel assemblies, which may significantly affect the maximum fuel temperature. In this paper, a three-dimensional computational fluid dynamics (CFD) analysis has been carried out on a typical fuel assembly of a prismatic VHTR. Thermal behaviours and heat transfer within the fuel assembly are intensively investigated using the CFD solutions. In addition, the accuracy of the unit cell approach is assessed against the CFD solutions. Two example situations are illustrated to demonstrate the deficiency of the unit cell model caused by neglecting the effects of the bypass gap flow and the radial power distribution within the fuel assembly

Enhanced thermal properties of novel shape-stabilized PEG composite phase change materials with radial mesoporous silica sphere for thermal energy storage

OpenAIRE

Min, Xin; Fang, Minghao; Huang, Zhaohui; Liu, Yan?gai; Huang, Yaoting; Wen, Ruilong; Qian, Tingting; Wu, Xiaowen

2015-01-01

Radial mesoporous silica (RMS) sphere was tailor-made for further applications in producing shape-stabilized composite phase change materials (ss-CPCMs) through a facile self-assembly process using CTAB as the main template and TEOS as SiO2 precursor. Novel ss-CPCMs composed of polyethylene glycol (PEG) and RMS were prepared through vacuum impregnating method. Various techniques were employed to characterize the structural and thermal properties of the ss-CPCMs. The DSC results indicated that...

PULSTAR fuel, low enrichment, long lifetime, economical, proven

International Nuclear Information System (INIS)

Carter, Robert E.; Leonard, Bobby E.

1993-01-01

In 1962, the Western New York Research Center, Inc., located at the State University of New York at Buffalo, decided they had a need for a reactor with pulsing and high power steady state capabilities. Both General Atomic and the American Machine and Foundry Corporation (AMF) were contacted to ascertain if it were feasible to construct a dual purpose reactor of this type. The General Atomic proposal indicated the feasibility but would not warrant a steady state power of 2 MW with ultimate capability of 5 MW. AMF did provide a conceptual design for such a dual reactor, call the PULSTAR, and sufficient design information to confirm that the operating specifications could be met. The PULSTAR fuel consisted of 6 enrichment UO 2 sintered pellets in zircaloy tubes (pins) mounted in a x 5 array inside a fuel assembly. The fuel design was patterned after fuel that was under development for light water power reactors and that had been extensively tested under high power pulse conditions in the SPERT Test Reactor. The fuel assemblies are rectangular in a horizontal cross section, 315 inches by 2.74 inches, allowing for flat control blades to be inserted in the core grid arrangement. The active height of the core is approximately 24 inches. In the initial Buffalo AMF contract, a collaborative development agreement was signed in conjunction with agreement to construct the facility. After completion of the Buffalo PULSTAR Reactor, the PULSTAR fuel underwent an extensive test program which resulted in some minor changes in the basic design. In 1965, North Carolina State University contracted with AMF for the construction of a dual MW steady state (with ultimate capability of 5 MW and pulsing PULSTAR Research Reactor. Their fuel is identical to the Buffalo fuel except for having an enrichment of 4% U-235. This paper presented basic information about the characteristics and performance of the PULSTAR Research Reactor fuel. The following summarizes this information. The fuel is of

PULSTAR fuel, low enrichment, long lifetime, economical, proven

Energy Technology Data Exchange (ETDEWEB)

Carter, Robert E; Leonard, Bobby E [Institute for Resource Management, Inc., Bethesda, MD (United States)

1993-08-01

In 1962, the Western New York Research Center, Inc., located at the State University of New York at Buffalo, decided they had a need for a reactor with pulsing and high power steady state capabilities. Both General Atomic and the American Machine and Foundry Corporation (AMF) were contacted to ascertain if it were feasible to construct a dual purpose reactor of this type. The General Atomic proposal indicated the feasibility but would not warrant a steady state power of 2 MW with ultimate capability of 5 MW. AMF did provide a conceptual design for such a dual reactor, call the PULSTAR, and sufficient design information to confirm that the operating specifications could be met. The PULSTAR fuel consisted of 6 enrichment UO{sub 2} sintered pellets in zircaloy tubes (pins) mounted in a x 5 array inside a fuel assembly. The fuel design was patterned after fuel that was under development for light water power reactors and that had been extensively tested under high power pulse conditions in the SPERT Test Reactor. The fuel assemblies are rectangular in a horizontal cross section, 315 inches by 2.74 inches, allowing for flat control blades to be inserted in the core grid arrangement. The active height of the core is approximately 24 inches. In the initial Buffalo AMF contract, a collaborative development agreement was signed in conjunction with agreement to construct the facility. After completion of the Buffalo PULSTAR Reactor, the PULSTAR fuel underwent an extensive test program which resulted in some minor changes in the basic design. In 1965, North Carolina State University contracted with AMF for the construction of a dual MW steady state (with ultimate capability of 5 MW and pulsing PULSTAR Research Reactor. Their fuel is identical to the Buffalo fuel except for having an enrichment of 4% U-235. This paper presented basic information about the characteristics and performance of the PULSTAR Research Reactor fuel. The following summarizes this information. The fuel

Three invariant Hi-C interaction patterns: Applications to genome assembly.

Science.gov (United States)

Oddes, Sivan; Zelig, Aviv; Kaplan, Noam

2018-06-01

Assembly of reference-quality genomes from next-generation sequencing data is a key challenge in genomics. Recently, we and others have shown that Hi-C data can be used to address several outstanding challenges in the field of genome assembly. This principle has since been developed in academia and industry, and has been used in the assembly of several major genomes. In this paper, we explore the central principles underlying Hi-C-based assembly approaches, by quantitatively defining and characterizing three invariant Hi-C interaction patterns on which these approaches can build: Intrachromosomal interaction enrichment, distance-dependent interaction decay and local interaction smoothness. Specifically, we evaluate to what degree each invariant pattern holds on a single locus level in different species, cell types and Hi-C map resolutions. We find that these patterns are generally consistent across species and cell types but are affected by sequencing depth, and that matrix balancing improves consistency of loci with all three invariant patterns. Finally, we overview current Hi-C-based assembly approaches in light of these invariant patterns and demonstrate how local interaction smoothness can be used to easily detect scaffolding errors in extremely sparse Hi-C maps. We suggest that simultaneously considering all three invariant patterns may lead to better Hi-C-based genome assembly methods. Copyright © 2018 Elsevier Inc. All rights reserved.

Star-shaped tetrathiafulvalene oligomers towards the construction of conducting supramolecular assembly.

Science.gov (United States)

Iyoda, Masahiko; Hasegawa, Masashi

2015-01-01

The construction of redox-active supramolecular assemblies based on star-shaped and radially expanded tetrathiafulvalene (TTF) oligomers with divergent and extended conjugation is summarized. Star-shaped TTF oligomers easily self-aggregate with a nanophase separation to produce supramolecular structures, and their TTF units stack face-to-face to form columnar structures using the fastener effect. Based on redox-active self-organizing supramolecular structures, conducting nanoobjects are constructed by doping of TTF oligomers with oxidants after the formation of such nanostructures. Although radical cations derived from TTF oligomers strongly interact in solution to produce a mixed-valence dimer and π-dimer, it seems to be difficult to produce nanoobjects of radical cations different from those of neutral TTF oligomers. In some cases, however, radical cations form nanostructured fibers and rods by controlling the supramolecular assembly, oxidation states, and counter anions employed.

Galaxy and Mass Assembly (GAMA): halo formation times and halo assembly bias on the cosmic web

Science.gov (United States)

Tojeiro, Rita; Eardley, Elizabeth; Peacock, John A.; Norberg, Peder; Alpaslan, Mehmet; Driver, Simon P.; Henriques, Bruno; Hopkins, Andrew M.; Kafle, Prajwal R.; Robotham, Aaron S. G.; Thomas, Peter; Tonini, Chiara; Wild, Vivienne

2017-09-01

We present evidence for halo assembly bias as a function of geometric environment (GE). By classifying Galaxy and Mass Assembly (GAMA) galaxy groups as residing in voids, sheets, filaments or knots using a tidal tensor method, we find that low-mass haloes that reside in knots are older than haloes of the same mass that reside in voids. This result provides direct support to theories that link strong halo tidal interactions with halo assembly times. The trend with GE is reversed at large halo mass, with haloes in knots being younger than haloes of the same mass in voids. We find a clear signal of halo downsizing - more massive haloes host galaxies that assembled their stars earlier. This overall trend holds independently of GE. We support our analysis with an in-depth exploration of the L-Galaxies semi-analytic model, used here to correlate several galaxy properties with three different definitions of halo formation time. We find a complex relationship between halo formation time and galaxy properties, with significant scatter. We confirm that stellar mass to halo mass ratio, specific star formation rate (SFR) and mass-weighed age are reasonable proxies of halo formation time, especially at low halo masses. Instantaneous SFR is a poor indicator at all halo masses. Using the same semi-analytic model, we create mock spectral observations using complex star formation and chemical enrichment histories, which approximately mimic GAMA's typical signal-to-noise ratio and wavelength range. We use these mocks to assert how well potential proxies of halo formation time may be recovered from GAMA-like spectroscopic data.

Radial electric fields for improved tokamak performance

International Nuclear Information System (INIS)

Downum, W.B.

1981-01-01

The influence of externally-imposed radial electric fields on the fusion energy output, energy multiplication, and alpha-particle ash build-up in a TFTR-sized, fusing tokamak plasma is explored. In an idealized tokamak plasma, an externally-imposed radial electric field leads to plasma rotation, but no charge current flows across the magnetic fields. However, a realistically-low neutral density profile generates a non-zero cross-field conductivity and the species dependence of this conductivity allows the electric field to selectively alter radial particle transport

Large-scale systematic analysis of 2D fingerprint methods and parameters to improve virtual screening enrichments.

Science.gov (United States)

Sastry, Madhavi; Lowrie, Jeffrey F; Dixon, Steven L; Sherman, Woody

2010-05-24

A systematic virtual screening study on 11 pharmaceutically relevant targets has been conducted to investigate the interrelation between 8 two-dimensional (2D) fingerprinting methods, 13 atom-typing schemes, 13 bit scaling rules, and 12 similarity metrics using the new cheminformatics package Canvas. In total, 157 872 virtual screens were performed to assess the ability of each combination of parameters to identify actives in a database screen. In general, fingerprint methods, such as MOLPRINT2D, Radial, and Dendritic that encode information about local environment beyond simple linear paths outperformed other fingerprint methods. Atom-typing schemes with more specific information, such as Daylight, Mol2, and Carhart were generally superior to more generic atom-typing schemes. Enrichment factors across all targets were improved considerably with the best settings, although no single set of parameters performed optimally on all targets. The size of the addressable bit space for the fingerprints was also explored, and it was found to have a substantial impact on enrichments. Small bit spaces, such as 1024, resulted in many collisions and in a significant degradation in enrichments compared to larger bit spaces that avoid collisions.

Uranium enrichment

International Nuclear Information System (INIS)

Rae, H.K.; Melvin, J.G.

1988-06-01

Canada is the world's largest producer and exporter of uranium, most of which is enriched elsewhere for use as fuel in LWRs. The feasibility of a Canadian uranium-enrichment enterprise is therefore a perennial question. Recent developments in uranium-enrichment technology, and their likely impacts on separative work supply and demand, suggest an opportunity window for Canadian entry into this international market. The Canadian opportunity results from three particular impacts of the new technologies: 1) the bulk of the world's uranium-enrichment capacity is in gaseous diffusion plants which, because of their large requirements for electricity (more than 2000 kW·h per SWU), are vulnerable to competition from the new processes; 2) the decline in enrichment costs increases the economic incentive for the use of slightly-enriched uranium (SEU) fuel in CANDU reactors, thus creating a potential Canadian market; and 3) the new processes allow economic operation on a much smaller scale, which drastically reduces the investment required for market entry and is comparable with the potential Canadian SEU requirement. The opportunity is not open-ended. By the end of the century the enrichment supply industry will have adapted to the new processes and long-term customer/supplier relationships will have been established. In order to seize the opportunity, Canada must become a credible supplier during this century

Interactions between Radial Electric Field, Transport and Structure in Helical Plasmas

International Nuclear Information System (INIS)

Ida, Katsumi and others

2006-01-01

Control of the radial electric field is considered to be important in helical plasmas, because the radial electric field and its shear are expected to reduce neoclassical and anomalous transport, respectively. Particle and heat transport, that determines the radial structure of density and electron profiles, sensitive to the structure of radial electric field. On the other hand, the radial electric field itself is determined by the plasma parameters. In general, the sign of the radial electric field is determined by the plasma collisionality, while the magnitude of the radial electric field is determined by the temperature and/or density gradients. Therefore the structure of radial electric field and temperature and density are strongly coupled through the particle and heat transport and formation mechanism of radial electric field. Interactions between radial electric field, transport and structure in helical plasmas is discussed based on the experiments on Large Helical Device

Anomalous Medial Branch of Radial Artery: A Rare Variant

Directory of Open Access Journals (Sweden)

Surbhi Wadhwa

2016-10-01

Full Text Available Radial artery is an important consistent vessel of the upper limb. It is a useful vascular access site for coronary procedures and its reliable anatomy has resulted in an elevation of radial forearm flaps for reconstructive surgeries of head and neck. Technical failures, in both the procedures, are mainly due to anatomical variations, such as radial loops, ectopic radial arteries or tortuosity in the vessel. We present a rare and a unique anomalous medial branch of the radial artery spiraling around the flexor carpi radialis muscle in the forearm with a high rising superficial palmar branch of radial artery. Developmentally it probably is a remanent of the normal pattern of capillary vessel maintenance and regression. Such a case is of importance for reconstructive surgeons and coronary interventionists, especially in view of its unique medial and deep course.

Neutronic characteristics of linear-assembly breed-and-burn reactors

International Nuclear Information System (INIS)

Petroski, Robert; Forget, Benoit; Forsberg, Charles

2012-01-01

Highlights: ► Simple models used to characterize general behavior of linear-assembly B and B reactors. ► Diffusion theory model developed to explain axial distributions, height vs. reactivity. ► Neutron excess concept reformulated to include linear-assembly B and B reactors. ► Designed model of B and B reactor started using melt-refined B and B reactor used fuel. ► Computed doubling time of fuel cycle requiring no chemical separations. - Abstract: Linear-assembly breed-and-burn (B and B) reactors are B and B reactors that use axially connected assemblies similar to conventional LWR or fast reactor fuel assemblies. Methods for analyzing linear-assembly B and B reactors and their fuel cycles are developed and applied. General neutronic characteristics of linear-assembly B and B reactors are analyzed, including the effects that burnup, shuffling sequence, and radial and axial size have on equilibrium-cycle k-effective. The mechanisms that give rise to a highly peaked axial burnup distribution are explained, and a method for predicting peak burnup vs. k-effective based on infinite-medium depletion calculations is developed. Next, the neutron excess concept from previous studies of B and B reactors is extended to apply to linear-assembly B and B reactors, which allows the amount of starter fuel needed to establish a given equilibrium cycle to be calculated. Several example applications of the neutron excess formulation are given. First, an example model of a linear-assembly B and B reactor is analyzed to find the neutron excess cost of an equilibrium cycle. Second, simple one-dimensional models are used to predict the neutron excess value obtainable from different starter fuel configurations. Finally, these ideas are applied to design a fuel cycle consisting of linear-assembly B and B reactors and fuel recycling via a melt refining process. The neutron excess concept is used to design an appropriate starter fuel configuration made from melt refined fuel, which

An axially and radially two-zoned large liquid-metal fast breeder reactor core concept

International Nuclear Information System (INIS)

Kamei, T.; Arie, K.; Moriki, Y.; Suzuki, M.; Yamaoka, M.

1985-01-01

A new core concept that has advantages over conventional homogeneous cores in neutronics characteristics such as power peaking factor, burnup reactivity loss, and reactivity response to the movement of control rods in earthquakes has been evolved. Two options of the new core concept are feasible. One is the so-called axially heterogeneous core, with the internal blanket placed at the lower part of the core. The other concept is similar to the conventional homogeneous core, but has two different plutonium-enriched zones in the axial as well as in the radial direction, so it is a hybrid type of the conventional homogeneous core and the axially heterogeneous core. The new design concept is described and the way that the core characteristics are improved by the chosen key parameters is shown

Clinical and Radiographic Outcomes of Unipolar and Bipolar Radial Head Prosthesis in Patients with Radial Head Fracture: A Systemic Review and Meta-Analysis.

Science.gov (United States)

Chen, Hongwei; Wang, Ziyang; Shang, Yongjun

2018-06-01

To compare clinical outcomes of unipolar and bipolar radial head prosthesis in the treatment of patients with radial head fracture. Medline, Cochrane, EMBASE, Google Scholar databases were searched until April 18, 2016 using the following search terms: radial head fracture, elbow fracture, radial head arthroplasty, implants, prosthesis, unipolar, bipolar, cemented, and press-fit. Randomized controlled trials, retrospective, and cohort studies were included. The Mayo elbow performance score (MEPS), disabilities of the arm, shoulder, and hand (DASH) score, radiologic assessment, ROM, and grip strength following elbow replacement were similar between prosthetic devices. The pooled mean excellent/good ranking of MEPS was 0.78 for unipolar and 0.73 for bipolar radial head arthroplasty, and the pooled mean MEPS was 86.9 and 79.9, respectively. DASH scores for unipolar and bipolar prosthesis were 19.0 and 16.3, respectively. Range of motion outcomes were similar between groups, with both groups have comparable risk of flexion arc, flexion, extension deficit, rotation arc, pronation, and supination (p values bipolar prosthesis). However, bipolar radial head prosthesis was associated with an increased chance of heterotopic ossification and lucency (p values ≤0.049) while unipolar prosthesis was not (p values ≥0.088). Both groups had risk for development of capitellar osteopenia or erosion/wear (p values ≤0.039). Unipolar and bipolar radial head prostheses were similar with respect to clinical outcomes. Additional comparative studies are necessary to further compare different radial head prostheses used to treat radial head fracture.

Criticality Calculations of Fresh LEU and MOX Assemblies for Transport and Storage at the Balakovo Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Goluoglu, S.

2001-01-11

Transportation of low-enriched uranium (LEU) and mixed-oxide (MOX) assemblies to and within the VVER-1000-type Balakovo Nuclear Power Plant is investigated. Effective multiplication factors for fresh fuel assemblies on the railroad platform, fresh fuel assemblies in the fuel transportation vehicle, and fresh fuel assemblies in the spent fuel storage pool are calculated. If there is no absorber between the units, the configurations with all MOX assemblies result in higher effective multiplication factors than the configurations with all LEU assemblies when the system is dry. When the system is flooded, the configurations with all LEU assemblies result in higher effective multiplication factors. For normal operating conditions, effective multiplication factors for all configurations are below the presumed upper subcritical limit of 0.95. For an accident condition of a fully loaded fuel transportation vehicle that is flooded with low-density water (possibly from a fire suppression system), the presumed upper subcritical limit is exceeded by configurations containing LEU assemblies.

Vitreous veils and radial lattice in Marshall syndrome.

Science.gov (United States)

Brubaker, Jacob W; Mohney, Brian G; Pulido, Jose S; Babovic-Vuksanovic, Dusica

2008-12-01

To report the findings of membranous vitreous veils and radial lattice in a child with Marshall syndrome. Observational case report. Retrospective review of medical records and fundus photograph of a 6-year-old boy with Marshall syndrome. Vitreoretinal findings were significant for bilateral membranous vitreous veils and radial lattice degeneration. This case demonstrates the occurrence of vitreous veils and radial lattice degeneration in patients with Marshall syndrome.

A segmented, enriched N-type germanium detector for neutrinoless double beta-decay experiments

Science.gov (United States)

Leviner, L. E.; Aalseth, C. E.; Ahmed, M. W.; Avignone, F. T.; Back, H. O.; Barabash, A. S.; Boswell, M.; De Braeckeleer, L.; Brudanin, V. B.; Chan, Y.-D.; Egorov, V. G.; Elliott, S. R.; Gehman, V. M.; Hossbach, T. W.; Kephart, J. D.; Kidd, M. F.; Konovalov, S. I.; Lesko, K. T.; Li, Jingyi; Mei, D.-M.; Mikhailov, S.; Miley, H.; Radford, D. C.; Reeves, J.; Sandukovsky, V. G.; Umatov, V. I.; Underwood, T. A.; Tornow, W.; Wu, Y. K.; Young, A. R.

2014-01-01

We present data characterizing the performance of the first segmented, N-type Ge detector, isotopically enriched to 85% 76Ge. This detector, based on the Ortec PT6×2 design and referred to as SEGA (Segmented, Enriched Germanium Assembly), was developed as a possible prototype for neutrinoless double beta-decay measurements by the MAJORANA collaboration. We present some of the general characteristics (including bias potential, efficiency, leakage current, and integral cross-talk) for this detector in its temporary cryostat. We also present an analysis of the resolution of the detector, and demonstrate that for all but two segments there is at least one channel that reaches the MAJORANA resolution goal below 4 keV FWHM at 2039 keV, and all channels are below 4.5 keV FWHM.

Long-Term Follow-up of Modular Metallic Radial Head Replacement: Commentary on an article by Jonathan P. Marsh, MD, FRCSC, et al.: "Radial Head Fractures Treated with Modular Metallic Radial Head Replacement: Outcomes at a Mean Follow-up of Eight Years".

OpenAIRE

Mansat, Pierre

2016-01-01

Radial head arthroplasty is used to stabilize the joint after a complex acute radial head fracture that is not amenable for fixation or to treat sequelae of radial head fractures. Most of the currently used radial head prostheses are metallic monoblock implants that are not consistently adaptable and raise technical challenges since their implantation requires lateral elbow subluxation. Metallic modular radial head arthroplasty implants available in various head and stem sizes have been devel...

First assembly phase for the ATLAS toroid coils

CERN Document Server

Patrice Loïez

2003-01-01

The ATLAS barrel toroid system consists of eight coils, each of axial length 25.3 m, assembled radially and symmetrically around the beam axis. The coils are of a flat racetrack type with two double-pancake windings made of 20.5 kA aluminium-stabilized niobium-titanium superconductor. In the first phase of assembly, the two 'pancakes' are packed into their vacuum vessel. This is done using bladders filled with resin and glass microbeads under pressure. The resin is heated and, once cooled, holds the pancakes in place. The operation has to be performed on both sides of the coil, which necessitated a special technique to turn the coils over and then transport them to the heating table. Photos 01, 02, 03: Use of the overhead travelling crane to hoist the coil up and then tilt it over, the coil frame's metal feet being used as rotational pivots, supporting half the coil's weight. Once it has been turned over, the coil, now with only half the frame, is transported to the heating table using a special lifting gant...

Advanced enrichment techniques

International Nuclear Information System (INIS)

Johnson, A.

1988-01-01

BNFL is in a unique position in that it has commercial experience of diffusion enrichment, and of centrifuge enrichment through its associate company Urenco. In addition BNFL is developing laser enrichment techniques as part of a UK development programme in this area. The paper describes the development programme which led to the introduction of competitive centrifuge enrichment technology by Urenco and discusses the areas where improvements have and will continue to be made in the centrifuge process. It also describes the laser development programme currently being undertaken in the UK. The paper concludes by discussing the relative merits of the various methods of uranium enrichment, with particular reference to the enrichment market likely to obtain over the rest of the century

Advanced enrichment techniques

International Nuclear Information System (INIS)

Johnson, A.

1987-01-01

BNFL is in a unique position in that it has commercial experience of diffusion enrichment, and of centrifuge enrichment through its associate company Urenco. In addition BNFL is developing laser enrichment techniques as part of a UK development programme in this area. The paper describes the development programme which led to the introduction of competitive centrifuge enrichment technology by Urenco and discusses the areas where improvements have and will continue to be made in the centrifuge process. It also describes the laser development programme currently being undertaken in the UK. The paper concludes by discussing the relative merits of the various methods of uranium enrichment, with particular reference to the enrichment market likely to obtain over the rest of the century. (author)

Central Reactivity Measurements on Assemblies 1 and 3 of the Fast Reactor FR0

International Nuclear Information System (INIS)

Londen, S.O.

1966-01-01

The reactivity effects of small samples of various materials have been measured, by the period method at the core centre of Assemblies 1 and 3 of the fast zero power reactor FR0. For some materials the reactivity change as a function of sample size has also been determined experimentally. The core of Assembly 1 consisted only of uranium enriched to 20 % whereas the core of Assembly 3 was diluted with 30 % graphite. The results have been compared with calculated values obtained with a second-order transport-theoretical perturbation model and using differently shielded cross sections depending upon sample size. Qualitative agreement has generally been found, although discrepancies still exist. The spectrum perturbation caused by the experimental arrangement has been analyzed and found to be rather important

Central Reactivity Measurements on Assemblies 1 and 3 of the Fast Reactor FR0

Energy Technology Data Exchange (ETDEWEB)

Londen, S O

1966-01-15

The reactivity effects of small samples of various materials have been measured, by the period method at the core centre of Assemblies 1 and 3 of the fast zero power reactor FR0. For some materials the reactivity change as a function of sample size has also been determined experimentally. The core of Assembly 1 consisted only of uranium enriched to 20 % whereas the core of Assembly 3 was diluted with 30 % graphite. The results have been compared with calculated values obtained with a second-order transport-theoretical perturbation model and using differently shielded cross sections depending upon sample size. Qualitative agreement has generally been found, although discrepancies still exist. The spectrum perturbation caused by the experimental arrangement has been analyzed and found to be rather important.

Buildup of radioxenon isotopes in MOX-assemblies

Energy Technology Data Exchange (ETDEWEB)

Gniffke, Thomas; Kirchner, Gerald [Carl Friedrich von Weizsaecker-Centre for Science and Peace Research, Hamburg (Germany)

2015-07-01

Radioxenon is the main tracer for detection of nuclear tests conducted underground under the verification regime of the Comprehensive Nuclear Test Ban Treaty (CTBT). Since radioxenon is emitted by civilian sources too, like commercial nuclear reactors, source discrimination is still an important issue. Inventory calculations are necessary to predict which xenon isotopic ratios are built up in a reactor and how they differ from those generated by a nuclear explosion. The screening line actually used by the CTBT Organization for source discrimination is based on calculations for uranium fuel of various enrichments used in pressurized water reactors (PWRs). The usage of different fuel, especially mixed U/Pu oxide (MOX) assemblies with reprocessed plutonium, may alter the radioxenon signature of civilian reactors. In this talk, calculations of the radioxenon buildup in a MOX-assembly used in a commercial PWR are presented. Implications for the CTBT verification regimes are discussed and open questions are addressed.

Juvenile psittacine environmental enrichment.

Science.gov (United States)

Simone-Freilicher, Elisabeth; Rupley, Agnes E

2015-05-01

Environmental enrichment is of great import to the emotional, intellectual, and physical development of the juvenile psittacine and their success in the human home environment. Five major types of enrichment include social, occupational, physical, sensory, and nutritional. Occupational enrichment includes exercise and psychological enrichment. Physical enrichment includes the cage and accessories and the external home environment. Sensory enrichment may be visual, auditory, tactile, olfactory, or taste oriented. Nutritional enrichment includes variations in appearance, type, and frequency of diet, and treats, novelty, and foraging. Two phases of the preadult period deserve special enrichment considerations: the development of autonomy and puberty. Copyright © 2015 Elsevier Inc. All rights reserved.

Radial head fracture associated with posterior interosseous nerve injury

Directory of Open Access Journals (Sweden)

Bernardo Barcellos Terra

Full Text Available ABSTRACT Fractures of the radial head and radial neck correspond to 1.7-5.4% of all fractures and approximately 30% may present associated injuries. In the literature, there are few reports of radial head fracture with posterior interosseous nerve injury. This study aimed to report a case of radial head fracture associated with posterior interosseous nerve injury. CASE REPORT: A male patient, aged 42 years, sought medical care after falling from a skateboard. The patient related pain and limitation of movement in the right elbow and difficulty to extend the fingers of the right hand. During physical examination, thumb and fingers extension deficit was observed. The wrist extension showed a slight radial deviation. After imaging, it became evident that the patient had a fracture of the radial head that was classified as grade III in the Mason classification. The patient underwent fracture fixation; at the first postoperative day, thumb and fingers extension was observed. Although rare, posterior interosseous nerve branch injury may be associated with radial head fractures. In the present case, the authors believe that neuropraxia occurred as a result of the fracture hematoma and edema.

Gas centrifuge enrichment plants inspection frequency and remote monitoring issues for advanced safeguards implementation

International Nuclear Information System (INIS)

Boyer, Brian David; Erpenbeck, Heather H.; Miller, Karen A.; Ianakiev, Kiril D.; Reimold, Benjamin A.; Ward, Steven L.; Howell, John

2010-01-01

Current safeguards approaches used by the IAEA at gas centrifuge enrichment plants (GCEPs) need enhancement in order to verify declared low enriched uranium (LEU) production, detect undeclared LEU production and detect high enriched uranium (BEU) production with adequate probability using non destructive assay (NDA) techniques. At present inspectors use attended systems, systems needing the presence of an inspector for operation, during inspections to verify the mass and 235 U enrichment of declared cylinders of uranium hexafluoride that are used in the process of enrichment at GCEPs. This paper contains an analysis of how possible improvements in unattended and attended NDA systems including process monitoring and possible on-site destructive analysis (DA) of samples could reduce the uncertainty of the inspector's measurements providing more effective and efficient IAEA GCEPs safeguards. We have also studied a few advanced safeguards systems that could be assembled for unattended operation and the level of performance needed from these systems to provide more effective safeguards. The analysis also considers how short notice random inspections, unannounced inspections (UIs), and the concept of information-driven inspections can affect probability of detection of the diversion of nuclear material when coupled to new GCEPs safeguards regimes augmented with unattended systems. We also explore the effects of system failures and operator tampering on meeting safeguards goals for quantity and timeliness and the measures needed to recover from such failures and anomalies.

Uranium enrichment

International Nuclear Information System (INIS)

1989-01-01

GAO was asked to address several questions concerning a number of proposed uranium enrichment bills introduced during the 100th Congress. The bill would have restructured the Department of Energy's uranium enrichment program as a government corporation to allow it to compete more effectively in the domestic and international markets. Some of GAO's findings discussed are: uranium market experts believe and existing market models show that the proposed DOE purchase of a $750 million of uranium from domestic producers may not significantly increase production because of large producer-held inventories; excess uranium enrichment production capacity exists throughout the world; therefore, foreign producers are expected to compete heavily in the United States throughout the 1990s as utilities' contracts with DOE expire; and according to a 1988 agreement between DOE's Offices of Nuclear Energy and Defense Programs, enrichment decommissioning costs, estimated to total $3.6 billion for planning purposes, will be shared by the commercial enrichment program and the government

Radial Field Piezoelectric Diaphragms

Science.gov (United States)

Bryant, R. G.; Effinger, R. T., IV; Copeland, B. M., Jr.

2002-01-01

A series of active piezoelectric diaphragms were fabricated and patterned with several geometrically defined Inter-Circulating Electrodes "ICE" and Interdigitated Ring Electrodes "ICE". When a voltage potential is applied to the electrodes, the result is a radially distributed electric field that mechanically strains the piezoceramic along the Z-axis (perpendicular to the applied electric field). Unlike other piezoelectric bender actuators, these Radial Field Diaphragms (RFDs) strain concentrically yet afford high displacements (several times that of the equivalent Unimorph) while maintaining a constant circumference. One of the more intriguing aspects is that the radial strain field reverses itself along the radius of the RFD while the tangential strain remains relatively constant. The result is a Z-deflection that has a conical profile. This paper covers the fabrication and characterization of the 5 cm. (2 in.) diaphragms as a function of poling field strength, ceramic thickness, electrode type and line spacing, as well as the surface topography, the resulting strain field and displacement as a function of applied voltage at low frequencies. The unique features of these RFDs include the ability to be clamped about their perimeter with little or no change in displacement, the environmentally insulated packaging, and a highly repeatable fabrication process that uses commodity materials.

Fiber optical assembly for fluorescence spectrometry

Science.gov (United States)

Carpenter, II, Robert W.; Rubenstein, Richard; Piltch, Martin; Gray, Perry

2010-12-07

A system for analyzing a sample for the presence of an analyte in a sample. The system includes a sample holder for containing the sample; an excitation source, such as a laser, and at least one linear array radially disposed about the sample holder. Radiation from the excitation source is directed to the sample, and the radiation induces fluorescent light in the sample. Each linear array includes a plurality of fused silica optical fibers that receive the fluorescent light and transmits a fluorescent light signal from the first end to an optical end port of the linear array. An end port assembly having a photo-detector is optically coupled to the optical end port. The photo-detector detects the fluorescent light signal and converts the fluorescent light signal into an electrical signal.

Isotope enrichment

International Nuclear Information System (INIS)

Lydtin, H-J.; Wilden, R.J.; Severin, P.J.W.

1978-01-01

The isotope enrichment method described is based on the recognition that, owing to mass diffusion and thermal diffusion in the conversion of substances at a heated substrate while depositing an element or compound onto the substrate, enrichment of the element, or a compound of the element, with a lighter isotope will occur. The cycle is repeated for as many times as is necessary to obtain the degree of enrichment required

Radial transport with perturbed magnetic field

Energy Technology Data Exchange (ETDEWEB)

Hazeltine, R. D. [Institute for Fusion Studies, University of Texas at Austin, Austin, Texas 78712 (United States)

2015-05-15

It is pointed out that the viscosity coefficient describing radial transport of toroidal angular momentum is proportional to the second power of the gyro-radius—like the corresponding coefficients for particle and heat transport—regardless of any geometrical symmetry. The observation is widely appreciated, but worth emphasizing because some literature gives the misleading impression that asymmetry can allow radial moment transport in first-order.

Radial transport with perturbed magnetic field

International Nuclear Information System (INIS)

Hazeltine, R. D.

2015-01-01

It is pointed out that the viscosity coefficient describing radial transport of toroidal angular momentum is proportional to the second power of the gyro-radius—like the corresponding coefficients for particle and heat transport—regardless of any geometrical symmetry. The observation is widely appreciated, but worth emphasizing because some literature gives the misleading impression that asymmetry can allow radial moment transport in first-order

Multi-recycling of transuranic elements in a PWR assembly with reduced fuel rod diameter

Energy Technology Data Exchange (ETDEWEB)

Chambers, Alex, E-mail: acchamb@gmail.com; Ragusa, Jean C., E-mail: jean.ragusa@tamu.edu

2014-04-01

Highlights: • Study of multiple recycling passes of transuranic elements: (a) without exceeding 5 wt.% on U-235 enrichment; (b) using PWR fuel assemblies compatible with current reactor core internals. • Isotopic concentrations tend towards an equilibrium after 15 recycle passes, suggesting that thermal recycling may be continued beyond that point. • Radiotoxicity comparisons for once-through UOX, once-recycle MOX-Pu, and multiple recycle passes of MOX-PuNpAm and MOX-PuNpAmCm are presented. - Abstract: This paper examines the multi-recycling of transuranic (TRU) elements (Pu-Np-Am-Cm) in standard Pressurized Water Reactor (PWR) assemblies. The original feed of TRU comes from legacy spent UOX fuel. For all subsequent recycling passes, TRU elements from the previous generation are employed, supplemented by TRU from legacy UOX fuel, as needed. The design criteria include: {sup 235}U enrichment requirements to remain below 5 w/o, TRU loading limits to avoid return to criticality under voided conditions, and assembly power peaking factors. In order to carry out multiple recycling passes within the design envelope, additional neutron moderation is required and achieved by reducing the fuel pellet diameter by about 13%, thus keeping the assembly design compatible with current PWR core internals. TRU transmutation rates and long-term ingestion radiotoxicity results are presented for 15 recycling passes and compared to standard UOX and MOX once-through cycles. The results also show that TRU fuel isotopics and radiotoxicity tend towards an equilibrium, enabling further additional recycling passes.

Experimental stress analysis of the attachment region of hemispherical shells with attached nozzles. Part 2b. Radial nozzle 7.875 in. O.D.--7.500 in. I.D. 10.00 in. penetration

International Nuclear Information System (INIS)

Maxwell, R.L.; Holland, R.W.; Stengl, G.R.

1970-06-01

The report presents the results of investigations conducted on a hemisphere with a radial nozzle of 7.875'' O.D. and 7.500'' I.D. and 10'' penetration into the hemisphere. Stress values were determined for the following five types of loadings: (1) internal pressure applied to the hemisphere and nozzle assembly, (2) an axial load applied collinear with nozzle, (3) a pure bending moment, or axial couple, applied to the nozzle, (4) a transverse or shear load applied normal to the nozzle, and (5) a pure torque applied in the radial plane of the nozzle

Self-assembled magnetic nanoparticle supported zeolitic imidazolate framework-8: An efficient adsorbent for the enrichment of triazine herbicides from fruit, vegetables, and water.

Science.gov (United States)

Zhou, Lian; Su, Ping; Deng, Yulan; Yang, Yi

2017-02-01

Zeolitic imidazolate frameworks have positive surface charges and high adsorption capabilities. In this work, zeolitic imidazolate frameworks-8 and negatively charged magnetic nanoparticles were self-assembled by electrostatic attraction under sonication. The extraction performance of the synthesized hybrid material was evaluated by using it as a magnetic adsorbent for the enrichment of triazine herbicides in various sample matrices prior to analysis using ultrafast liquid chromatography. The main parameters, that is, extraction time, adsorbent dosage, salt concentration, and desorption conditions, were evaluated. Under the optimum conditions, good linear responses from 2.5 to 200 ng/mL for atrazine (simazine) and 1 to 200 ng/mL for prometryn (ametryn), with correlation coefficients (R 2 ) higher than 0.9992 were obtained. The detection limits of the method (S/N = 3) were 0.18-0.72 ng/mL. The proposed method was successfully used to determine triazine herbicides in six samples, namely, apple, pear, strawberry, pakchoi, lettuce, and water. The amounts of simazine in all the fruit and vegetable samples were 10.8-25.2 ng/mL. The recoveries of all the analytes were 88.0-101.9%, with relative standard deviations of less than 8.8%. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Safeguards for reprocessing and enrichment plants

International Nuclear Information System (INIS)

1977-01-01

Agency safeguards are entering a new phase with the coming under active safeguards for the first time of reprocessing plants in several regions of the world. This is taking place at a time when not only the safeguards aspect itself is coming under international scrutiny, but also at a time when the necessity of reprocessing plants is being called into question. Attracting less attention at the moment, but potentially of equal significance, are the enrichment plants that soon will be coming under Agency safeguards. It is not unreasonable in view of the present controversies to ask what is the significance of these reprocessing and enrichment plants, what are the problems concerning safeguards that appear to have given rise to the controversies, and how these problems are to be solved. The question of significance is an easy one to answer. The output of these plants is material which some people consider can be used directly for military purposes, whereas the output from other plants, for instance, reactors, would require long and extensive processing before it could be used for military purposes. Like most short answers, this one is an over-simplification which requires some elaboration to make it strictly accurate. For example, the material output of a power reactor is in the form of irradiate assemblies containing plutonium which is potentially of military use if the irradiation had been within a certain range. However, to utilize this plutonium under clandestine conditions, the highly radioactive material would have to be secretly transported to a reprocessing plant and there would have to be simultaneous falsification of the reactor material accounts and the plant records. Such falsification would be difficult to conceal. The total time required to obtain usable plutonium would be many months. Diversion of material from a uranium fabrication plant making fuel for power reactors would be easier physically but strategically it would be of little value. The

Two dimensional, two fluid model for sodium boiling in LMFBR fuel assemblies

International Nuclear Information System (INIS)

Granziera, M.R.; Kazimi, M.S.

1980-05-01

A two dimensional numerical model for the simulation of sodium boiling transient was developed using the two fluid set of conservation equations. A semiimplicit numerical differencing scheme capable of handling the problems associated with the ill-posedness implied by the complex characteristic roots of the two fluid problems was used, which took advantage of the dumping effect of the exchange terms. Of particular interest in the development of the model was the identification of the numerical problems caused by the strong disparity between the axial and radial dimensions of fuel assemblies. A solution to this problem was found which uses the particular geometry of fuel assemblies to accelerate the convergence of the iterative technique used in the model. Three sodium boiling experiments were simulated with the model, with good agreement between the experimental results and the model predictions

Development of WWER-440 fuel. Use of fuel assemblies of 2-nd and 3-rd generations with increased enrichment

International Nuclear Information System (INIS)

Vasilchenko, I.; Lushin, V.; Ananev, U.; Baranov, A.; Kukushkin, U.

2009-01-01

The problem of increasing the power of units at NPPs with WWER-440 is of current importance. There are all the necessary prerequisites for the above-stated problem as a result of updating the design of fuel assemblies and codes. The decrease of power peaking factor in the core is achieved by using profiled fuel assemblies, fuel-integrated burning absorber, FAs with modernized docking unit, modern codes, which allows decreasing conservatism of RP safety substantiation. A wide range of experimental studies of fuel behaviour has been performed which has reached burn-up of (50-60) MW·day/kgU in transition and emergency conditions, post-reactor studies of fuel assemblies, fuel rods and fuel pellets with a 5-year operating period have been performed, which prove high reliability of fuel, presence of a large margin in the fuel pillar, which helps reactor operation at increased power. The results of the work performed on introduction of 5-6 fuel cycles show that the ultimate fuel state on operability in WWER-440 reactors is far from being achieved. Neutron-physical and thermal-hydraulic characteristics of the cores of working power units with RP V-213 are such that actual (design and measured) power peaking factors on fuel assemblies and fuel rods, as a rule, are smaller than the maximum design values. This factor is a real reserve for power forcing. There is experience of operating Units 1, 2, 4 of the Kola NPP and Unit 2 of the Rovno NPP at increased power. Units of the Loviisa NPP are operated at 109 % power. During transfer to work at increased power it is reasonable to use fuel assemblies with increased height of the fuel pillar, which allows decreasing medium linear power distribution. Further development of the 2-nd generation fuel assembly design and consequent transition to working fuel assemblies of the 3-rd generation provides significant improvement of fuel consumption under the conditions of WWER-440 reactors operation with more continuous fuel cycles and

Optimization of radially heterogeneous 1000-MW(e) LMFBR core configurations. Appendixes D and E. Research project 620-25

International Nuclear Information System (INIS)

Barthold, W.P.; Orechwa, Y.; Su, S.F.; Hutter, E.; Batch, R.V.; Beitel, J.C.; Turski, R.B.; Lam, P.S.K.

1979-11-01

A parameter study was conducted to determine the interrelated effects of: loosely or tightly coupled fuel regions separated by internal blanket assemblies, number of fuel regions, core height, number and arrangement of internal blanket subassemblies, number and size of fuel pins in a subassembly, etc. the effects of these parameters on sodium void reactivity, Doppler, incoherence, breeding gain, and thermohydraulics were of prime interest. Trends were established and ground work laid for optimization of a large, radially-heterogeneous, LMFBR core that will have low energetics in an HCDA and will have good thermal and breeding performance

New generation enrichment monitoring technology for gas centrifuge enrichment plants

International Nuclear Information System (INIS)

Ianakiev, Kiril D.; Alexandrov, Boian S.; Boyer, Brian D.; Hill, Thomas R.; Macarthur, Duncan W.; Marks, Thomas; Moss, Calvin E.; Sheppard, Gregory A.; Swinhoe, Martyn T.

2008-01-01

The continuous enrichment monitor, developed and fielded in the 1990s by the International Atomic Energy Agency, provided a go-no-go capability to distinguish between UF 6 containing low enriched (approximately 4% 235 U) and highly enriched (above 20% 235 U) uranium. This instrument used the 22-keV line from a 109 Cd source as a transmission source to achieve a high sensitivity to the UF 6 gas absorption. The 1.27-yr half-life required that the source be periodically replaced and the instrument recalibrated. The instrument's functionality and accuracy were limited by the fact that measured gas density and gas pressure were treated as confidential facility information. The modern safeguarding of a gas centrifuge enrichment plant producing low-enriched UF 6 product aims toward a more quantitative flow and enrichment monitoring concept that sets new standards for accuracy stability, and confidence. An instrument must be accurate enough to detect the diversion of a significant quantity of material, have virtually zero false alarms, and protect the operator's proprietary process information. We discuss a new concept for advanced gas enrichment assay measurement technology. This design concept eliminates the need for the periodic replacement of a radioactive source as well as the need for maintenance by experts. Some initial experimental results will be presented.

Concepts of radial and angular kinetic energies

DEFF Research Database (Denmark)

Dahl, Jens Peder; Schleich, W.P.

2002-01-01

We consider a general central-field system in D dimensions and show that the division of the kinetic energy into radial and angular parts proceeds differently in the wave-function picture and the Weyl-Wigner phase-space picture, Thus, the radial and angular kinetic energies are different quantities...

Radial gradient and radial deviation radiomic features from pre-surgical CT scans are associated with survival among lung adenocarcinoma patients.

Science.gov (United States)

Tunali, Ilke; Stringfield, Olya; Guvenis, Albert; Wang, Hua; Liu, Ying; Balagurunathan, Yoganand; Lambin, Philippe; Gillies, Robert J; Schabath, Matthew B

2017-11-10

The goal of this study was to extract features from radial deviation and radial gradient maps which were derived from thoracic CT scans of patients diagnosed with lung adenocarcinoma and assess whether these features are associated with overall survival. We used two independent cohorts from different institutions for training (n= 61) and test (n= 47) and focused our analyses on features that were non-redundant and highly reproducible. To reduce the number of features and covariates into a single parsimonious model, a backward elimination approach was applied. Out of 48 features that were extracted, 31 were eliminated because they were not reproducible or were redundant. We considered 17 features for statistical analysis and identified a final model containing the two most highly informative features that were associated with lung cancer survival. One of the two features, radial deviation outside-border separation standard deviation, was replicated in a test cohort exhibiting a statistically significant association with lung cancer survival (multivariable hazard ratio = 0.40; 95% confidence interval 0.17-0.97). Additionally, we explored the biological underpinnings of these features and found radial gradient and radial deviation image features were significantly associated with semantic radiological features.

Illumination Profile & Dispersion Variation Effects on Radial Velocity Measurements

Science.gov (United States)

Grieves, Nolan; Ge, Jian; Thomas, Neil B.; Ma, Bo; Li, Rui; SDSS-III

2015-01-01

The Multi-object APO Radial-Velocity Exoplanet Large-Area Survey (MARVELS) measures radial velocities using a fiber-fed dispersed fixed-delay interferometer (DFDI) with a moderate dispersion spectrograph. This setup allows a unique insight into the 2D illumination profile from the fiber on to the dispersion grating. Illumination profile investigations show large changes in the profile over time and fiber location. These profile changes are correlated with dispersion changes and long-term radial velocity offsets, a major problem within the MARVELS radial velocity data. Characterizing illumination profiles creates a method to both detect and correct radial velocity offsets, allowing for better planet detection. Here we report our early results from this study including improvement of radial velocity data points from detected giant planet candidates. We also report an illumination profile experiment conducted at the Kitt Peak National Observatory using the EXPERT instrument, which has a DFDI mode similar to MARVELS. Using profile controlling octagonal-shaped fibers, long term offsets over a 3 month time period were reduced from ~50 m/s to within the photon limit of ~4 m/s.

Radial transfer effects for poloidal rotation

Science.gov (United States)

Hallatschek, Klaus

2010-11-01

Radial transfer of energy or momentum is the principal agent responsible for radial structures of Geodesic Acoustic Modes (GAMs) or stationary Zonal Flows (ZF) generated by the turbulence. For the GAM, following a physical approach, it is possible to find useful expressions for the individual components of the Poynting flux or radial group velocity allowing predictions where a mathematical full analysis is unfeasible. Striking differences between up-down symmetric flux surfaces and asymmetric ones have been found. For divertor geometries, e.g., the direction of the propagation depends on the sign of the ion grad-B drift with respect to the X-point, reminiscent of a sensitive determinant of the H-mode threshold. In nonlocal turbulence computations it becomes obvious that the linear energy transfer terms can be completely overwhelmed by the action of the turbulence. In contrast, stationary ZFs are governed by the turbulent radial transfer of momentum. For sufficiently large systems, the Reynolds stress becomes a deterministic functional of the flows, which can be empirically determined from the stress response in computational turbulence studies. The functional allows predictions even on flow/turbulence states not readily obtainable from small amplitude noise, such as certain transport bifurcations or meta-stable states.

Low enrichment Mo-99 target development program at ANSTO

International Nuclear Information System (INIS)

Donlevy, Therese M.; Anderson, Peter J.; Beattie, David; Braddock, Ben; Fulton, Scott; Godfrey, Robert; Law, Russell; McNiven, Scott; Sirkka, Pertti; Storr, Greg; Wassink, David; Wong, Alan; Yeoh, Guan

2002-01-01

The Australian Nuclear Science and Technology Organisation (ANSTO, formerly AAEC) has been producing fission product Mo-99 in HIFAR, from the irradiation of Low Enrichment Uranium (LEU) UO 2 targets, for nearly thirty years. Over this period, the U-235 enrichment has been increased in stages, from natural to 1.8% to 2.2%. The decision to provide Australia with a replacement research reactor (RRR) for HIFAR has created an ideal opportunity to review and improve the current Mo-99 production process from target design through to chemical processing and waste management options. ANSTO has entered into a collaboration with Argonne National Laboratory (RERTR) to develop a target using uranium metal foil with U-235 enrichment of less than 20% The initial focus has been to demonstrate use of LEU foil targets in HIFAR, using existing irradiation methodology. The current effort focussed on designing a target assembly with optimised thermohydraulic characteristics to accommodate larger LEU foils to meet Mo-99 production needs. The ultimate goal is to produce an LEU target suitable for use in the Replacement Research Reactor when it is commissioned in 2005. This paper reports our activities on: - The regulatory approval processes required in order to undertake irradiation of this new target; -Supporting calculations (neutronics, computational fluid dynamics) for safety submission; - Design challenges and changes to prototype irradiation; - Trial irradiation of LEU foil target in HIFAR; - Future target and rig development program at ANSTO. (author)

Anterior transposition of the radial nerve--a cadaveric study.

Science.gov (United States)

Yakkanti, Madhusudhan R; Roberts, Craig S; Murphy, Joshua; Acland, Robert D

2008-01-01

The radial nerve is at risk during the posterior plating of the humerus. The purpose of this anatomic study was to assess the extent of radial nerve dissection required for anterior transposition through the fracture site (transfracture anterior transposition). A cadaver study was conducted approaching the humerus by a posterior midline incision. The extent of dissection of the nerve necessary for plate fixation of the humerus fracture was measured. An osteotomy was created to model a humeral shaft fracture at the spiral groove (OTA classification 12-A2, 12-A3). The radial nerve was then transposed anterior to the humeral shaft through the fracture site. The additional dissection of the radial nerve and the extent of release of soft tissue from the humerus shaft to achieve the transposition were measured. Plating required a dissection of the radial nerve 1.78 cm proximal and 2.13 cm distal to the spiral groove. Transfracture anterior transposition of the radial nerve required an average dissection of 2.24 cm proximal and 2.68 cm distal to the spiral groove. The lateral intermuscular septum had to be released for 2.21 cm on the distal fragment to maintain laxity of the transposed nerve. Transfracture anterior transposition of the radial nerve before plating is feasible with dissection proximal and distal to the spiral groove and elevation of the lateral intermuscular septum. Potential clinical advantages of this technique include enhanced fracture site visualization, application of broader plates, and protection of the radial nerve during the internal fixation.

Uranium enrichment plans

International Nuclear Information System (INIS)

Thomas, D.C.; Gagne, R.W.

1978-01-01

The following topics are covered: the status of the Government's existing uranium enrichment services contracts, natural uranium requirements based on the latest contract information, uncertainty in predicting natural uranium requirements based on uranium enrichment contracts, and domestic and foreign demand assumed in enrichment planning

A New Filtering Algorithm Utilizing Radial Velocity Measurement

Institute of Scientific and Technical Information of China (English)

LIU Yan-feng; DU Zi-cheng; PAN Quan

2005-01-01

Pulse Doppler radar measurements consist of range, azimuth, elevation and radial velocity. Most of the radar tracking algorithms in engineering only utilize position measurement. The extended Kalman filter with radial velocity measureneut is presented, then a new filtering algorithm utilizing radial velocity measurement is proposed to improve tracking results and the theoretical analysis is also given. Simulation results of the new algorithm, converted measurement Kalman filter, extended Kalman filter are compared. The effectiveness of the new algorithm is verified by simulation results.

Advanced PWR Core Design with Siemens High-Plutonium-Content MOX Fuel Assemblies

International Nuclear Information System (INIS)

Dieter Porsch; Gerhard Schlosser; Hans-Dieter Berger

2000-01-01

The Siemens experience with plutonium recycling dates back to the late 1960s. Over the years, extensive research and development programs were performed for the qualification of mixed-oxide (MOX) technology and design methods. Today's typical reload enrichments for uranium and MOX fuel assemblies and modern core designs have become more demanding with respect to accuracy and reliability of design codes. This paper presents the status of plutonium recycling in operating high-burnup pressurized water reactor (PWR) cores. Based on actual examples, it describes the validation status of the design methods and stresses current and future needs for fuel assembly and core design including those related to the disposition of weapons-grade plutonium

Photoelectric Radial Velocities, Paper XIX Additional Spectroscopic ...

Indian Academy of Sciences (India)

ian velocity curve that does justice to the measurements, but it cannot be expected to have much predictive power. Key words. Stars: late-type—stars: radial velocities—spectroscopic binaries—orbits. 0. Preamble. The 'Redman K stars' are a lot of seventh-magnitude K stars whose radial velocities were first observed by ...

Radioactive isotope production for medical applications using Kharkov electron driven subcritical assembly facility

International Nuclear Information System (INIS)

Talamo, A.; Gohar, Y.

2007-01-01

Kharkov Institute of Physics and Technology (KIPT) of Ukraine has a plan to construct an accelerator driven subcritical assembly. The main functions of the subcritical assembly are the medical isotope production, neutron thereby, and the support of the Ukraine nuclear industry. Reactor physics experiments and material research will be carried out using the capabilities of this facility. The United States of America and Ukraine have started collaboration activity for developing a conceptual design for this facility with low enrichment uranium (LEU) fuel. Different conceptual designs are being developed based on the facility mission and the engineering requirements including nuclear physics, neutronics, heat transfer, thermal hydraulics, structure, and material issues. Different fuel designs with LEU and reflector materials are considered in the design process. Safety, reliability, and environmental considerations are included in the facility conceptual design. The facility is configured to accommodate future design improvements and upgrades. This report is a part of the Argonne National Laboratory Activity within this collaboration for developing and characterizing the subcritical assembly conceptual design. In this study, the medical isotope production function of the Kharkov facility is defined. First, a review was carried out to identify the medical isotopes and its medical use. Then a preliminary assessment was performed without including the self-shielding effect of the irradiated samples. Finally, more detailed investigation was carried out including the self-shielding effect, which defined the sample size and irradiation location for producing each medical isotope. In the first part, the reaction rates were calculated as the multiplication of the cross section with the unperturbed neutron flux of the facility. Over fifty isotopes were considered and all transmutation channels are used including (n,γ), (n,2n), (n,p), and (γ,n). In the second part, the parent

Radioactive isotope production for medical applications using Kharkov electron driven subcritical assembly facility.

Energy Technology Data Exchange (ETDEWEB)

Talamo, A.; Gohar, Y.; Nuclear Engineering Division

2007-05-15

Kharkov Institute of Physics and Technology (KIPT) of Ukraine has a plan to construct an accelerator driven subcritical assembly. The main functions of the subcritical assembly are the medical isotope production, neutron thereby, and the support of the Ukraine nuclear industry. Reactor physics experiments and material research will be carried out using the capabilities of this facility. The United States of America and Ukraine have started collaboration activity for developing a conceptual design for this facility with low enrichment uranium (LEU) fuel. Different conceptual designs are being developed based on the facility mission and the engineering requirements including nuclear physics, neutronics, heat transfer, thermal hydraulics, structure, and material issues. Different fuel designs with LEU and reflector materials are considered in the design process. Safety, reliability, and environmental considerations are included in the facility conceptual design. The facility is configured to accommodate future design improvements and upgrades. This report is a part of the Argonne National Laboratory Activity within this collaboration for developing and characterizing the subcritical assembly conceptual design. In this study, the medical isotope production function of the Kharkov facility is defined. First, a review was carried out to identify the medical isotopes and its medical use. Then a preliminary assessment was performed without including the self-shielding effect of the irradiated samples. Finally, more detailed investigation was carried out including the self-shielding effect, which defined the sample size and irradiation location for producing each medical isotope. In the first part, the reaction rates were calculated as the multiplication of the cross section with the unperturbed neutron flux of the facility. Over fifty isotopes were considered and all transmutation channels are used including (n,{gamma}), (n,2n), (n,p), and ({gamma},n). In the second part

Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

International Nuclear Information System (INIS)

Chodak, P. III

1996-05-01

This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO 2 assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the 239 Pu and ≥90% total Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products

Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

Energy Technology Data Exchange (ETDEWEB)

Chodak, III, Paul [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

1996-05-01

This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO₂ assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the ²³⁹Pu and ≥90% {sub total}Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products.

Other enrichment related contracts

International Nuclear Information System (INIS)

Hall, J.C.

1978-01-01

In addition to long-term enrichment contracts, DOE has other types of contracts: (1) short-term, fixed-commitment enrichment contract; (2) emergency sales agreement for enriched uranium; (3) feed material lease agreement; (4) enriched uranium storage agreement; and (5) feed material usage agreement

Linear theory radial and nonradial pulsations of DA dwarf stars

International Nuclear Information System (INIS)

Starrfield, S.; Cox, A.N.; Hodson, S.; Pesnell, W.D.

1982-01-01

The Los Alamos stellar envelope and radial linear non-adiabatic computer code, along with a new Los Alamos non-radial code are used to investigate the total hydrogen mass necessary to produce the non-radial instability of DA dwarfs

Role of Gag and lipids during HIV-1 assembly in CD4 T cells and Macrophages

Directory of Open Access Journals (Sweden)

Charlotte eMariani

2014-06-01

Full Text Available HIV-1 is an RNA enveloped virus that preferentiallyinfects CD4+ T lymphocytes andalso macrophages. In CD4+ T cells, HIV-1mainly buds from the host cell plasma membrane.The viral Gag polyprotein targets theplasma membrane and is the orchestrator ofthe HIV assembly as its expression is sufficientto promote the formation of virus-likeparticles particles carrying a lipidic envelopederiving from the host cell membrane. Certainlipids are enriched in the viral membraneand are thought to play a key role in theassembly process and the envelop composition.A large body of work performed oninfected CD4+ T cells has provided importantknowledge about the assembly process andthe membrane virus lipid composition. WhileHIV assembly and budding in macrophages isthought to follow the same general Gag-drivenmechanism as in T-lymphocytes, the HIV cyclein macrophage exhibits specific features.In these cells, new virions bud from the limitingmembrane of seemingly intracellular compartments,where they accumulate while remaininginfectious. These structures are now oftenreferred to as Virus Containing Compartments(VCCs. Recent studies suggest that VCCsrepresent intracellularly sequestered regionsof the plasma membrane, but their precisenature remains elusive. The proteomic andlipidomic characterization of virions producedby T cells or macrophages has highlightedthe similarity between their composition andthat of the plasma membrane of producercells, as well as their enrichment in acidiclipids, some components of raft lipids andin tetraspanin-enriched microdomains. Greatchances are that Gag promotes the coalescenceof these components into an assemblyplatform from which viral budding takesplace. How Gag exactly interacts with membranelipids and what are the mechanisms involvedin the interaction between the differentmembrane nanodomains within the assemblyplatform remains unclear. Here we review recentliterature regarding the role of Gag andlipids

Validation of the TUBRNP model with the radial distribution of plutonium in MOX fuel measured by SIMS and EPMA

Energy Technology Data Exchange (ETDEWEB)

O` Carroll, C; Laar, J Van De; Walker, C T [CEC Joint Research Centre, Karlsruhe (Germany)

1997-08-01

The new model TUBRNP (TRANSURANUS burnup) predicts the radial power density distribution as a function of burnup (and hence the radial burnup profile as a function of time) together with the radial profile of plutonium. Comparisons between measurements and the prediction of the TUBRNP model have been made for UO{sub 2} LWR fuels: they were found to be in excellent agreement and it is seen that TUBRNP is a marked improved on previous models. A powerful techniques for the characterization of irradiation fuel is Electron Probe Microanalysis (EPMA). Uranium, plutonium and fission product distributions can be analysed quantitatively. A complement, providing isotopic information with a lateral resolution comparable to EPMA, is secondary ion mass spectrometry (SIMS). Recently, the technique has been successfully applied for the measurement of the radial distribution of plutonium isotopes in irradiated nuclear fuel pins. The extension of the TUBRNP model to mixed oxide fuels seems to be the natural step to take. In MOX fuels the picture is greatly complicated by the presence of the (U, Pu)O{sub 2} agglomerates. The rim effect referred to above may be masked by the high concentrations of plutonium in the bulk of the fuel. A detailed investigation of a number of MOX fuel samples has been made using the TUBRNP model. Results are presented for a range of fuels with different enrichment and burnup. Through its participation in the PRIMO and DOMO programmes, PSI in conjunction with the Institute for Transuranium Elements had the opportunity to validate the new theoretical model TUBRNP. The authors with therefore to express their thanks to the organizers and to the numerous European and Japanese organizations which have supported these two international programmes on MOX fuel behavior. 7 refs, 9 figs, 3 tabs.

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

Calculation of Savannah River K Reactor Mark-22 assembly LOCA/ECS power limits

International Nuclear Information System (INIS)

Fischer, S.R.; Farman, R.F.; Birdsell, S.A.

1992-01-01

This paper summarizes the results of TRAC-PF1/MOD3 calculations of Mark-22 fuel assembly of loss-of-coolant accident/emergency cooling system (LOCA/ECS) power limits for the Savannah River Site (SRS) K Reactor. This effort was part of a larger effort undertaken by the Los Alamos National Laboratory for the US Department of Energy to perform confirmatory power limits calculations for the SRS K Reactor. A method using a detailed three-dimensional (3D) TRAC model of the Mark-22 fuel assembly was developed to compute LOCA/ECS power limits. Assembly power was limited to ensure that no point on the fuel assembly walls would exceed the local saturation temperature. The detailed TRAC model for the Mark-22 assembly consisted of three concentric 3D vessel components which simulated the two targets, two fuel tubes, and three main flow channels of the fuel assembly. The model included 100% eccentricity between the assembly annuli and a 20% power tilt. Eccentricity in the radial alignment of the assembly annuli arises because axial spacer ribs that run the length of the fuel and targets are used. Wall-shear, interfacial-shear, and wall heat-transfer correlations were developed and implemented in TRAC-PF1/MOD3 specifically for modeling flow and heat transfer in the narrow ribbed annuli encountered in the Mark-22 fuel assembly design. We established the validity of these new constitutive models using separate-effects benchmarks. TRAC system calculations of K Reactor indicated that the limiting ECS-phase accident is a double-ended guillonite break in a process water line at the pump discharge (i.e., a PDLOCA). The fuel assembly with the minimum cooling potential is identified from this system calculation. Detailed assembly calculations then were performed using appropriate boundary conditions obtained from this limiting system LOCA. Coolant flow rates and pressure boundary conditions were obtained from this system calculation and applied to the detailed assembly model

Passive gamma analysis of the boiling-water-reactor assemblies

Energy Technology Data Exchange (ETDEWEB)

Vo, D., E-mail: ducvo@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM (United States); Favalli, A. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Schwalbach, P. [European Atomic Energy Community (EURATOM), Luxemburg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Tobin, S.; Trellue, H. [Los Alamos National Laboratory, Los Alamos, NM (United States); Vaccaro, S. [European Atomic Energy Community (EURATOM), Luxemburg (Luxembourg)

2016-09-11

This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden's Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative–Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: {sup 137}Cs, {sup 154}Eu, {sup 134}Cs, and to a lesser extent, {sup 106}Ru and {sup 144}Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.

Effect of the radial electric field on turbulence

International Nuclear Information System (INIS)

Carreras, B.A.; Lynch, V.E.

1990-01-01

For many years, the neoclassical transport theory for three- dimensional magnetic configurations, such as magnetic mirrors, ELMO Bumpy Tori (EBTs), and stellarators, has recognized the critical role of the radial electric field in the confinement. It was in these confinement devices that the first experimental measurements of the radial electric field were made and correlated with confinement losses. In tokamaks, the axisymmetry implies that the neoclassical fluxes are ambipolar and, as a consequence, independent of the radial electric field. However, axisymmetry is not strict in a tokamak with turbulent fluctuations, and near the limiter ambipolarity clearly breaks down. Therefore, the question of the effect of the radial electric field on tokamak confinement has been raised in recent years. In particular, the radial electric field has been proposed to explain the transition from L-mode to H-mode confinement. There is some initial experimental evidence supporting this type of explanation, although there is not yet a self-consistent theory explaining the generation of the electric field and its effect on the transport. Here, a brief review of recent results is presented. 27 refs., 4 figs

A user's evaluation of radial flow HEPA filters

International Nuclear Information System (INIS)

Purcell, J.A.

1992-07-01

High efficiency particulate air (HEPA) filters of rectangular cross section have been used to remove particulates and the associated radioactivity from air ventilation streams since the advent of nuclear materials processing. Use of round axial flow HEPA filters is also longstanding. The advantages of radial flow filters in a circular configuration have been well demonstrated in UKAEA during the last 5--7 years. An evaluation of radial flow filters for fissile process gloveboxes reveals several substantial benefits in addition to the advantages claimed in UKAEA Facilities. The radial flow filter may be provided in a favorable geometry resulting in improved criticality safety. The filter configuration lends to in-place testing at the glovebox to exhaust duct interface. This will achieve compliance with DOE Order 6430.1A, Section 99.0.2. Preliminary testing at SRS for radial flow filters manufactured by Flanders Filters, Inc. revealed compliance in all the usual specifications for filtration efficiency, pressure differential and materials of construction. An evaluation, further detailed in this report, indicates that the radial flow HEPA filter should be considered for inclusion in new ventilation system designs

Radial extension of drift waves in presence of velocity profiles

International Nuclear Information System (INIS)

Sen, S.; Weiland, J.

1994-01-01

The effect of a radially varying poloidal velocity field on the recently found radially extended toroidal drift waves is investigated analytically. The role of velocity curvature (υ φ '') is found to have robust effects on the radial model structure of the mode. For a positive value of the curvature (Usually found in the H-mode edges) the radial model envelope, similar to the sheared slab case, becomes fully outgoing. The mode is therefore stable. On the other hand, for a negative value of the curvature (usually observed in the L-mode edges) all the characteristics of conventional drift waves return back. The radial mode envelope reduces to a localized Gaussian shape and the mode is therefore unstable again for typical (magnetic) shear values in tokamaks. Velocity shear (υ φ ??) on the other hand is found to have rather insignificant role both in determining the radial model structure and stability

Combined Radial and Femoral Access Strategy and Radial-Femoral Rendezvous in Patients With Long and Complex Iliac Occlusions.

Science.gov (United States)

Hanna, Elias B; Mogabgab, Owen N; Baydoun, Hassan

2018-01-01

We present cases of complex, calcified iliac occlusive disease revascularized via a combined radial-femoral access strategy. Through a 6-French, 125-cm transradial guiding catheter, antegrade guidewires and catheters are advanced into the iliac occlusion, while retrograde devices are advanced transfemorally. The transradial and transfemoral channels communicate, allowing the devices to cross the occlusion into the true lumen (radial-femoral antegrade-retrograde rendezvous).

A feasibility study for the application of enriched gadolinia burnable absorber rods in nuclear core design

International Nuclear Information System (INIS)

Lee, Chung Chan; Zee, Sung Quun; Kim, Kang Seog; Song, Jae Seung

2000-12-01

An analysis model using MICBURN-3/CASMO-3 is established for the enriched gadolinia burnable absorber rods. A homogenized cross section editing code, PROLOG, is modified so that it can handle such a fuel assembly that includes two different types of gadolinia rods. Study shows that Gd-155 and Gd-157 are almost same in suppressing the excess reactivity and it is recommended to enrich both odd number isotopes, Gd-155 and Gd-157. It is estimated that the cycle length increases by 2 days if enriched gadolinia rods are used in the commercial nuclear power plant such as YGN-3 of which the cycle length is assumed 2 years. For the advanced integral reactor SMART in which ultra long cycle length and soluble boron-free operation concept is applied, natural gadolinia burnable absorber rods fail to control the excess reactivity. On the other hand, enriched gadolinia rods are successful in controling the excess reactivity. To minimize power peakings, various placements of gadolinia rods are tested. Also initial reactivity holddown and gadolinia burnout time are parametrized with respect to the number of gadolinia rods and gadolinia weight fractions

On-Line Enrichment Monitor for UF{sub 6} Gas Centrifuge Enrichment Plant

Energy Technology Data Exchange (ETDEWEB)

Ianakiev, K. D.; Boyer, B.; Favalli, A.; Goda, J. M.; Hill, T.; Keller, C.; Lombardi, M.; Paffett, M.; MacArthur, D. W.; McCluskey, C.; Moss, C. E.; Parker, R.; Smith, M. K.; Swinhoe, M. T. [Los Alamos National Laboratory, Los Alamos (United States)

2012-06-15

This paper is a continuation of the Advanced Enrichment Monitoring Technology for UF{sub 6} Gas Centrifuge Enrichment Plant (GCEP) work, presented in the 2010 IAEA Safeguards Symposium. Here we will present the system architecture for a planned side-by-side field trial test of passive (186-keV line spectroscopy and pressure-based correction for UF{sub 6} gas density) and active (186-keV line spectroscopy and transmission measurement based correction for UF{sub 6} gas density) enrichment monitoring systems in URENCO's enrichment plant in Capenhurst. Because the pressure and transmission measurements of UF{sub 6} are complementary, additional information on the importance of the presence of light gases and the UF{sub 6} gas temperature can be obtained by cross-correlation between simultaneous measurements of transmission, pressure and 186-keV intensity. We will discuss the calibration issues and performance in the context of accurate, on-line enrichment measurement. It is hoped that a simple and accurate on-line enrichment monitor can be built using the UF{sub 6} gas pressure provided by the Operator, based on online mass spectrometer calibration, assuming a negligible (a small fraction of percent) contribution of wall deposits. Unaccounted-for wall deposits present at the initial calibration will lead to unwanted sensitivity to changes in theUF{sub 6} gas pressure and thus to error in the enrichment results. Because the accumulated deposits in the cascade header pipe have been identified as an issue for Go/No Go measurements with the Cascade Header Enrichment Monitor (CHEM) and Continuous Enrichment Monitor (CEMO), it is important to explore their effect. Therefore we present the expected uncertainty on enrichment measurements obtained by propagating the errors introduced by deposits, gas density, etc. and will discuss the options for a deposit correction during initial calibration of an On-Line Enrichment Monitor (OLEM).

A novel structure of permanent-magnet-biased radial hybrid magnetic bearing

International Nuclear Information System (INIS)

Sun Jinji; Fang Jiancheng

2011-01-01

The paper proposes a novel structure for a permanent-magnet-biased radial hybrid magnetic bearing. Based on the air gap between the rotor and stator of traditional radial hybrid magnetic bearings, a subsidiary air gap is first constructed between the permanent magnets and the inner magnetic parts. Radial magnetic bearing makes X and Y magnetic fields independent of each other with separate stator poles, and the subsidiary air gap makes control flux to a close loop. As a result, magnetic field coupling of the X and Y channels is decreased significantly by the radial hybrid magnetic bearing and makes it easier to design control systems. Then an external rotor structure is designed into the radial hybrid magnetic bearing. The working principle of the radial hybrid magnetic bearing and its mathematical model is discussed. Finally, a non-linear magnetic network method is proposed to analyze the radial hybrid magnetic bearing. Simulation results indicate that magnetic fields in the two channels of the proposed radial hybrid magnetic bearing decouple well from each other.

A novel structure of permanent-magnet-biased radial hybrid magnetic bearing

Energy Technology Data Exchange (ETDEWEB)

Sun Jinji, E-mail: sunjinji@aspe.buaa.edu.c [Key Laboratory of Fundamental Science for National Defense, Novel Inertial Instrument and Navigation System Technology, School of Instrument Science and Opto-electronics Engineering, Beijing University of Aeronautics and Astronautics, 100191 (China); Fang Jiancheng [Key Laboratory of Fundamental Science for National Defense, Novel Inertial Instrument and Navigation System Technology, School of Instrument Science and Opto-electronics Engineering, Beijing University of Aeronautics and Astronautics, 100191 (China)

2011-01-15

The paper proposes a novel structure for a permanent-magnet-biased radial hybrid magnetic bearing. Based on the air gap between the rotor and stator of traditional radial hybrid magnetic bearings, a subsidiary air gap is first constructed between the permanent magnets and the inner magnetic parts. Radial magnetic bearing makes X and Y magnetic fields independent of each other with separate stator poles, and the subsidiary air gap makes control flux to a close loop. As a result, magnetic field coupling of the X and Y channels is decreased significantly by the radial hybrid magnetic bearing and makes it easier to design control systems. Then an external rotor structure is designed into the radial hybrid magnetic bearing. The working principle of the radial hybrid magnetic bearing and its mathematical model is discussed. Finally, a non-linear magnetic network method is proposed to analyze the radial hybrid magnetic bearing. Simulation results indicate that magnetic fields in the two channels of the proposed radial hybrid magnetic bearing decouple well from each other.

Detonation in supersonic radial outflow

KAUST Repository

Kasimov, Aslan R.

2014-11-07

We report on the structure and dynamics of gaseous detonation stabilized in a supersonic flow emanating radially from a central source. The steady-state solutions are computed and their range of existence is investigated. Two-dimensional simulations are carried out in order to explore the stability of the steady-state solutions. It is found that both collapsing and expanding two-dimensional cellular detonations exist. The latter can be stabilized by putting several rigid obstacles in the flow downstream of the steady-state sonic locus. The problem of initiation of standing detonation stabilized in the radial flow is also investigated numerically. © 2014 Cambridge University Press.

Radial nerve dysfunction (image)

Science.gov (United States)

The radial nerve travels down the arm and supplies movement to the triceps muscle at the back of the upper arm. ... the wrist and hand. The usual causes of nerve dysfunction are direct trauma, prolonged pressure on the ...

Neutron noise measurements at the Delphi subcritical assembly

International Nuclear Information System (INIS)

Szieberth, M.; Klujber, G.; Kloosterman, J. L.; De Haas, D.

2012-01-01

The paper presents the results and evaluations of a comprehensive set of neutron noise measurements on the Delphi subcritical assembly of the Delft Univ. of Technology. The measurements investigated the effect of different source distributions (inherent spontaneous fission and 252 Cf) and the position of the detectors applied (both radially and vertically). The evaluation of the measured data has been performed by the variance-to-mean ratio (VTMR, Feynman-α), the autocorrelation (ACF, Rossi-α) and the cross-correlation (CCF) methods. The values obtained for the prompt decay constant show a strong bias, which depends both on the detector position and on the source distribution. This is due to the presence of higher modes in the system. It has been observed that the α value fitted is higher when the detector is close to the boundary of the core or to the 252 Cf point-source. The higher alpha-modes have also been observed by fitting functions describing two alpha-modes. The successful set of measurement also provides a good basis for further theoretical investigations including the Monte Carlo simulation of the noise measurements and the calculation of the alpha-modes in the Delphi subcritical assembly. (authors)

Variable stator radial turbine

Science.gov (United States)

Rogo, C.; Hajek, T.; Chen, A. G.

1984-01-01

A radial turbine stage with a variable area nozzle was investigated. A high work capacity turbine design with a known high performance base was modified to accept a fixed vane stagger angle moveable sidewall nozzle. The nozzle area was varied by moving the forward and rearward sidewalls. Diffusing and accelerating rotor inlet ramps were evaluated in combinations with hub and shroud rotor exit rings. Performance of contoured sidewalls and the location of the sidewall split line with respect to the rotor inlet was compared to the baseline. Performance and rotor exit survey data are presented for 31 different geometries. Detail survey data at the nozzle exit are given in contour plot format for five configurations. A data base is provided for a variable geometry concept that is a viable alternative to the more common pivoted vane variable geometry radial turbine.

Radial-probe EBUS for the diagnosis of peripheral pulmonary lesions

Directory of Open Access Journals (Sweden)

Marcia Jacomelli

Full Text Available ABSTRACT Objective: Conventional bronchoscopy has a low diagnostic yield for peripheral pulmonary lesions. Radial-probe EBUS employs a rotating ultrasound transducer at the end of a probe that is passed through the working channel of the bronchoscope. Radial-probe EBUS facilitates the localization of peripheral pulmonary nodules, thus increasing the diagnostic yield. The objective of this study was to present our initial experience using radial-probe EBUS in the diagnosis of peripheral pulmonary lesions at a tertiary hospital. Methods: We conducted a retrospective analysis of 54 patients who underwent radial-probe EBUS-guided bronchoscopy for the investigation of pulmonary nodules or masses between February of 2012 and September of 2013. Radial-probe EBUS was performed with a flexible 20-MHz probe, which was passed through the working channel of the bronchoscope and advanced through the bronchus to the target lesion. For localization of the lesion and for collection procedures (bronchial brushing, transbronchial needle aspiration, and transbronchial biopsy, we used fluoroscopy. Results: Radial-probe EBUS identified 39 nodules (mean diameter, 1.9 ± 0.7 cm and 19 masses (mean diameter, 4.1 ± 0.9 cm. The overall sensitivity of the method was 66.7% (79.5% and 25.0%, respectively, for lesions that were visible and not visible by radial-probe EBUS. Among the lesions that were visible by radial-probe EBUS, the sensitivity was 91.7% for masses and 74.1% for nodules. The complications were pneumothorax (in 3.7% and bronchial bleeding, which was controlled bronchoscopically (in 9.3%. Conclusions: Radial-probe EBUS shows a good safety profile, a low complication rate, and high sensitivity for the diagnosis of peripheral pulmonary lesions.

ATLAS facility fabrication and assembly

CERN Document Server

Ballard, E O; Davis, H A; Nielsen, K E; Parker, G V; Parsons, W M

2001-01-01

Summary form only given. Atlas is a pulsed-power facility recently completed at Los Alamos National Laboratory to drive hydrodynamic experiments. This new generation pulsed-power machine consists of a radial array of 24, 240-kV Marx modules and transmission lines supplying current to the load region at the machine center. The transmission lines, powered by the Marx modules, consist of cable headers, load protection switches and tri-plates interfacing to the center transition section through detachable current joints. A conical power-flow-channel attaches to the transition section providing an elevated interface to attach the experimental loads for diagnostic access. Fabrication and assembly of all components for the Atlas machine was completed in August 2000. The machine has also progressed through a test phase where the Marx module/transmission line units were fired, individually, into a test load. Progression continued with eight and sixteen lines being fired. Subsequently, an overall machine test was condu...

Uranium Enrichment, an overview

International Nuclear Information System (INIS)

Coates, J.H.

1994-01-01

This general presentation on uranium enrichment will be followed by lectures on more specific topics including descriptions of enrichment processes and assessments of the prevailing commercial and industrial situations. I shall therefore avoid as much as possible duplications with these other lectures, and rather dwell on: some theoretical aspects of enrichment in general, underlying the differences between statistical and selective processes, a review and comparison between enrichment processes, remarks of general order regarding applications, the proliferation potential of enrichment. It is noteworthy that enrichment: may occur twice in the LWR fuel cycle: first by enriching natural uranium, second by reenriching uranium recovered from reprocessing, must meet LWR requirements, and in particular higher assays required by high burn up fuel elements, bears on the structure of the entire front part of the fuel cycle, namely in the conversion/reconversion steps only involving UF 6 for the moment. (author). tabs., figs., 4 refs

Sharp Dissection versus Electrocautery for Radial Artery Harvesting

Science.gov (United States)

Marzban, Mehrab; Arya, Reza; Mandegar, Mohammad Hossein; Karimi, Abbas Ali; Abbasi, Kiomars; Movahed, Namvar; Abbasi, Seyed Hesameddin

2006-01-01

Radial arteries have been increasingly used during the last decade as conduits for coronary artery revascularization. Although various harvesting techniques have been described, there has been little comparative study of arterial damage and patency. A radial artery graft was used in 44 consecutive patients, who were randomly divided into 2 groups. In the 1st group, the radial artery was harvested by sharp dissection and in the 2nd, by electrocautery. These groups were compared with regard to radial artery free flow, harvest time, number of clips used, complications, and endothelial damage. Radial artery free flow before and after intraluminal administration of papaverine was significantly greater in the electrocautery group (84.3 ± 50.7 mL/min and 109.7 ± 68.5 mL/min) than in the sharp-dissection group (52.9 ± 18.3 mL/min and 69.6 ± 28.2 mL/ min) (P =0.003). Harvesting time by electrocautery was significantly shorter (25.4 ± 4.3 min vs 34.4 ± 5.9 min) (P =0.0001). Electrocautery consumed an average of 9.76 clips, versus 22.45 clips consumed by sharp dissection. The 2 groups were not different regarding postoperative complications, except for 3 cases of temporary paresthesia of the thumb in the electrocautery group; histopathologic examination found no endothelial damage. We conclude that radial artery harvesting by electrocautery is faster and more economical than harvesting by sharp dissection and is associated with better intraoperative flow and good preservation of endothelial integrity. PMID:16572861

Radial distribution of ions in pores with a surface charge

NARCIS (Netherlands)

Stegen, J.H.G. van der; Görtzen, J.; Kuipers, J.A.M.; Hogendoorn, J.A.; Versteeg, G.F.

2001-01-01

A sorption model applicable to calculate the radial equilibrium concentrations of ions in the pores of ion-selective membranes with a pore structure is developed. The model is called the radial uptake model. Because the model is applied to a Nafion sulfonic layer with very small pores and the radial

The Matlab Radial Basis Function Toolbox

Directory of Open Access Journals (Sweden)

Scott A. Sarra

2017-03-01

Full Text Available Radial Basis Function (RBF methods are important tools for scattered data interpolation and for the solution of Partial Differential Equations in complexly shaped domains. The most straight forward approach used to evaluate the methods involves solving a linear system which is typically poorly conditioned. The Matlab Radial Basis Function toolbox features a regularization method for the ill-conditioned system, extended precision floating point arithmetic, and symmetry exploitation for the purpose of reducing flop counts of the associated numerical linear algebra algorithms.

Stellar Angular Momentum Distributions and Preferential Radial Migration

Science.gov (United States)

Wyse, Rosemary; Daniel, Kathryne J.

2018-04-01

I will present some results from our recent investigations into the efficiency of radial migration in stellar disks of differing angular momentum distributions, within a given adopted 2D spiral disk potential. We apply to our models an analytic criterion that determines whether or not individual stars are in orbits that could lead to radial migration around the corotation resonance. We couch our results in terms of the local stellar velocity dispersion and find that the fraction of stars that could migrate radially decreases as the velocity dispersion increases. I will discuss implications and comparisons with the results of other approaches.

Research on Radial Vibration of a Circular Plate

Directory of Open Access Journals (Sweden)

Wei Liu

2016-01-01

Full Text Available Radial vibration of the circular plate is presented using wave propagation approach and classical method containing Bessel solution and Hankel solution for calculating the natural frequency theoretically. In cylindrical coordinate system, in order to obtain natural frequency, propagation and reflection matrices are deduced at the boundaries of free-free, fixed-fixed, and fixed-free using wave propagation approach. Furthermore, radial phononic crystal is constructed by connecting two materials periodically for the analysis of band phenomenon. Also, Finite Element Simulation (FEM is adopted to verify the theoretical results. Finally, the radial and piezoelectric effects on the band are also discussed.

Radial-probe EBUS for the diagnosis of peripheral pulmonary lesions.

Science.gov (United States)

Jacomelli, Marcia; Demarzo, Sergio Eduardo; Cardoso, Paulo Francisco Guerreiro; Palomino, Addy Lidvina Mejia; Figueiredo, Viviane Rossi

2016-01-01

Conventional bronchoscopy has a low diagnostic yield for peripheral pulmonary lesions. Radial-probe EBUS employs a rotating ultrasound transducer at the end of a probe that is passed through the working channel of the bronchoscope. Radial-probe EBUS facilitates the localization of peripheral pulmonary nodules, thus increasing the diagnostic yield. The objective of this study was to present our initial experience using radial-probe EBUS in the diagnosis of peripheral pulmonary lesions at a tertiary hospital. We conducted a retrospective analysis of 54 patients who underwent radial-probe EBUS-guided bronchoscopy for the investigation of pulmonary nodules or masses between February of 2012 and September of 2013. Radial-probe EBUS was performed with a flexible 20-MHz probe, which was passed through the working channel of the bronchoscope and advanced through the bronchus to the target lesion. For localization of the lesion and for collection procedures (bronchial brushing, transbronchial needle aspiration, and transbronchial biopsy), we used fluoroscopy. Radial-probe EBUS identified 39 nodules (mean diameter, 1.9 ± 0.7 cm) and 19 masses (mean diameter, 4.1 ± 0.9 cm). The overall sensitivity of the method was 66.7% (79.5% and 25.0%, respectively, for lesions that were visible and not visible by radial-probe EBUS). Among the lesions that were visible by radial-probe EBUS, the sensitivity was 91.7% for masses and 74.1% for nodules. The complications were pneumothorax (in 3.7%) and bronchial bleeding, which was controlled bronchoscopically (in 9.3%). Radial-probe EBUS shows a good safety profile, a low complication rate, and high sensitivity for the diagnosis of peripheral pulmonary lesions. A broncoscopia convencional possui baixo rendimento diagnóstico para lesões pulmonares periféricas. A ecobroncoscopia radial (EBUS radial) emprega um transdutor ultrassonográfico rotatório na extremidade de uma sonda que é inserida no canal de trabalho do broncoscópio. O EBUS

Radial vibration and ultrasonic field of a long tubular ultrasonic radiator.

Science.gov (United States)

Shuyu, Lin; Zhiqiang, Fu; Xiaoli, Zhang; Yong, Wang; Jing, Hu

2013-09-01

The radial vibration of a metal long circular tube is studied analytically and its electro-mechanical equivalent circuit is obtained. Based on the equivalent circuit, the radial resonance frequency equation is derived. The theoretical relationship between the radial resonance frequency and the geometrical dimensions is studied. Finite element method is used to simulate the radial vibration and the radiated ultrasonic field and the results are compared with those from the analytical method. It is concluded that the radial resonance frequency for a solid metal rod is larger than that for a metal tube with the same outer radius. The radial resonance frequencies from the analytical method are in good agreement with those from the numerical method. Based on the acoustic field analysis, it is concluded that the long metal tube with small wall thickness is superior to that with large wall thickness in producing radial vibration and ultrasonic radiation. Therefore, it is expected to be used as an effective radial ultrasonic radiator in ultrasonic sewage treatment, ultrasonic antiscale and descaling and other ultrasonic liquid handling applications. Copyright © 2013 Elsevier B.V. All rights reserved.

Assembly and control of large microtubule complexes

Science.gov (United States)

Korolev, Kirill; Ishihara, Keisuke; Mitchison, Timothy

Motility, division, and other cellular processes require rapid assembly and disassembly of microtubule structures. We report a new mechanism for the formation of asters, radial microtubule complexes found in very large cells. The standard model of aster growth assumes elongation of a fixed number of microtubules originating from the centrosomes. However, aster morphology in this model does not scale with cell size, and we found evidence for microtubule nucleation away from centrosomes. By combining polymerization dynamics and auto-catalytic nucleation of microtubules, we developed a new biophysical model of aster growth. The model predicts an explosive transition from an aster with a steady-state radius to one that expands as a travelling wave. At the transition, microtubule density increases continuously, but aster growth rate discontinuously jumps to a nonzero value. We tested our model with biochemical perturbations in egg extract and confirmed main theoretical predictions including the jump in the growth rate. Our results show that asters can grow even though individual microtubules are short and unstable. The dynamic balance between microtubule collapse and nucleation could be a general framework for the assembly and control of large microtubule complexes. NIH GM39565; Simons Foundation 409704; Honjo International 486 Scholarship Foundation.

Feasibility study for LEU conversion of the WWR-K reactor at the Institute of Nuclear Physics in Kazakhstan using a 5-tube fuel assembly

International Nuclear Information System (INIS)

Hanan, N.A.; Liaw, J.R.; Matos, J.E.

2005-01-01

A feasibility study by the RERTR program for possible LEU conversion of the 6 MW WWR-K reactor concludes that conversion is feasible using an LEU 5-tube Russian fuel assembly design. This 5-tube design is one of several LEU fuel assembly designs being studied (Ref. 1) for possible use in this reactor. The 5-tube assembly contains 200 g 235 U with an enrichment of 19.7% in four cylindrical inner tubes and an outer hexagonal tube with the same external dimensions as the current HEU (36%) 5-tube fuel assembly, which contains 112.5 g 235 U. The fuel meat material, LEU UO 2 -Al dispersion fuel with ∼ 2.5 g U/cm 3 , has been extensively irradiation tested in a number of reactors with uranium enrichments of 36% and 19.7%. Since the 235 U loading of the LEU assemblies is much larger than the HEU assemblies, a smaller LEU core with five rows of fuel assemblies is possible (instead of six rows of fuel assemblies in the HEU core). This smaller LEU core would consume about 60% as many fuel assemblies per year as the current HEU core and provide thermal neutron fluxes in the inner irradiation channels that are ∼ 17% larger than with the present HEU core. The current 21 day cycle length would be maintained and the average discharge burnup would be ∼ 42%. Neutron fluxes in the five outer irradiation channels would be smaller in the LEU core unless these channels can be moved closer to the LEU fuel assemblies. Results show that the smaller LEU core would meet the reactor's shutdown margin requirements and would have an adequate thermal-hydraulic safety margin to onset of nucleate boiling. (author)

Criticality Analysis of SAMOP Subcritical Assembly

International Nuclear Information System (INIS)

Tegas-Sutondo; Syarip; Triwulan-Tjiptono

2005-01-01

A critically analysis has been performed for homogenous system of uranyl nitrate solution, as part of a preliminary design assessment on neutronic aspect of SAMOP sub-critical assembly. The analysis is intended to determine some critical parameters such as the minimum of critical dimension and critical mass for the desired concentration. As the basis of this analysis, it has been defined a fuel system with an enrichment of 20% for cylindrical geometry of both bare and graphite reflected of 30 cm thickness. The MCNP code has been utilized for this purpose, for variation of concentrations ranging from 150 g/l to 500 g/l. It is found that the best concentration giving the minimum geometrical dimension is around 400 g/l, for both the bare and reflected systems. Whilst the best one, of minimum critical mass is corresponding to the concentration of around 200 g/l with critical mass around 14.1 kg and 4.2 kg for the bare and reflected systems respectively. Based on the result of calculations, it is concluded that by taking into consideration of the critical limit, the SAMOP subcritical assembly is neutronically can be made. (author)

Radial Fuzzy Systems

Czech Academy of Sciences Publication Activity Database

Coufal, David

2017-01-01

Roč. 319, 15 July (2017), s. 1-27 ISSN 0165-0114 R&D Projects: GA MŠk(CZ) LD13002 Institutional support: RVO:67985807 Keywords : fuzzy systems * radial functions * coherence Subject RIV: BA - General Mathematics OBOR OECD: Computer sciences, information science, bioinformathics (hardware development to be 2.2, social aspect to be 5.8) Impact factor: 2.718, year: 2016

Acesso radial em intervenções coronarianas percutâneas: panorama atual brasileiro Acceso radial en intervenciones coronarias percutáneas: panorama actual brasileño Radial approach in percutaneous coronary interventions: current status in Brazil

Directory of Open Access Journals (Sweden)

Pedro Beraldo de Andrade

2011-04-01

Full Text Available FUNDAMENTO: Embora a técnica radial exiba resultados incontestáveis na redução de complicações vasculares e ocorrência de sangramento grave quando comparada à técnica femoral, seu emprego permanece restrito a poucos centros que a elegeram como via de acesso preferencial. OBJETIVO: Avaliar o cenário atual das intervenções coronarianas percutâneas no Brasil quanto à utilização da via de acesso radial. MÉTODOS: Análise dos dados cadastrados de forma espontânea na Central Nacional de Intervenções Cardiovasculares (CENIC durante o quadriênio de 2005-2008, o que totaliza 83.376 procedimentos. RESULTADOS: A técnica radial foi utilizada em 12,6% dos procedimentos efetivados, e a técnica femoral, em 84,3%. Os 3,1% restantes foram representados pela dissecção ou punção braquial. Com uma taxa de sucesso de 97,5%, a opção pelo acesso radial associou-se à redução significativa de complicações vasculares quando comparado ao femoral (2,5% versus 3,6%, p FUNDAMENTO: Aunque la técnica radial exhiba resultados incontestables en la reducción de complicaciones vasculares y ocurrencia de sangrado grave cuando es comparada a la técnica femoral, su empleo permanece restringido a pocos centros que la eligieron como vía de acceso preferencial. OBJETIVO:Evaluar el escenario actual de las intervenciones coronarias percutáneas en el Brasil en cuanto a la utilización de la vía de acceso radial. MÉTODOS:Análisis de los datos registrados de forma espontánea en la Central Nacional de Intervenciones Cardiovasculares (CENIC durante el cuatrienio de 2005-2008, lo que totaliza 83.376 procedimientos. RESULTADOS:La técnica radial fue utilizada en 12,6% de los procedimientos efectuados, y la técnica femoral, en 84,3%. Los 3,1% restantes fueron representados por la disección o punción braquial. Con una tasa de éxito de 97,5%, la opción por el acceso radial se asoció a la reducción significativa de complicaciones vasculares cuando

Optimization of Gad Pattern with Geometrical Weight

International Nuclear Information System (INIS)

Chang, Do Ik; Woo, Hae Seuk; Choi, Seong Min

2009-01-01

The prevailing burnable absorber for domestic nuclear power plants is a gad fuel rod which is used for the partial control of excess reactivity and power peaking. The radial peaking factor, which is one of the critical constraints for the plant safety depends largely on the number of gad bearing rods and the location of gad rods within fuel assembly. Also the concentration of gad, UO 2 enrichment in the gad fuel rod, and fuel lattice type play important roles for the resultant radial power peaking. Since fuel is upgraded periodically and longer fuel cycle management requires more burnable absorbers or higher gad weight percent, it is required frequently to search for the optimized gad patterns, i.e., the distribution of gad fuel rods within assembly, for the various fuel environment and fuel management changes. In this study, the gad pattern optimization algorithm with respect to radial power peaking factor using geometrical weight is proposed for a single gad weight percent, in which the candidates of the optimized gad pattern are determined based on the weighting of the gad rod location and the guide tube. Also the pattern evaluation is performed systematically to determine the optimal gad pattern for the various situation

Uranium enrichment

International Nuclear Information System (INIS)

Mohrhauer, H.

1982-01-01

The separation of uranium isotopes in order to enrich the fuel for light water reactors with the light isotope U-235 is an important part of the nuclear fuel cycle. After the basic principals of isotope separation the gaseous diffusion and the centrifuge process are explained. Both these techniques are employed on an industrial scale. In addition a short review is given on other enrichment techniques which have been demonstrated at least on a laboratory scale. After some remarks on the present situation on the enrichment market the progress in the development and the industrial exploitation of the gas centrifuge process by the trinational Urenco-Centec organisation is presented. (orig.)

United States uranium enrichment policies

International Nuclear Information System (INIS)

Roberts, R.W.

1977-01-01

ERDA's uranium enrichment program policies governing the manner in which ERDA's enrichment complex is being operated and expanded to meet customer requirements for separative work, research and development activities directed at providing technology alternatives for future enrichment capacity, and establishing the framework for additional domestic uranium enrichment capacity to meet the domestic and foreign nuclear industry's growing demand for enrichment services are considered. The ERDA enrichment complex consists of three gaseous diffusion plants located in Oak Ridge, Tennessee; Paducah, Kentucky; and Portsmouth, Ohio. Today, these plants provide uranium enrichment services for commercial nuclear power generation. These enrichment services are provided under contracts between the Government and the utility customers. ERDA's program involves a major pilot plant cascade, and pursues an advanced isotope separation technique for the late 1980's. That the United States must develop additional domestic uranium enrichment capacity is discussed

Atomics International fuel fabrication facility and low enrichment program [contributed by T.A. Moss, AI

International Nuclear Information System (INIS)

Moss, T.A.

1993-01-01

The AI facility is approximately 30,000 square feet in area and consists of four general areas. One area is devoted to the production of UAl x powder. It consists of a series of arc melting furnaces, crushing lines, glove boxes, and compacting presses. The second area is used for the rolling of fuel plates. The third area is used for the machining of the plates to final size and also the machining of the fuel elements. In the fourth area the fuel plates are swaged into assemblies, and all welding and inspection operations are performed. As part of the lower enrichment program we are scheduled to put a second UAl x powder line into operation and we have had to expand some of our storage area. Under the low enrichment program the AI fuel facility will be modified to accommodate a separate low enrichment Al x production line and compacting line. This facility modification should be done by the end of the fiscal year. We anticipate producing fuel with an enrichment slightly less than 20% We anticipate powder being available for plate production shortly after the facility is completed. Atomics International is scheduled to conduct plate LEU verification work using fully enriched material in the June-July time period, at which time we will investigate what level of uranium loadings we can go to using the current process. It is anticipated that 55 volume percent uranium compound in our fuel form can be achieved

Velocidades radiales en Collinder 121

Science.gov (United States)

Arnal, M.; Morrell, N.

Se han llevado a cabo observaciones espectroscópicas de unas treinta estrellas que son posibles miembros del cúmulo abierto Collinder 121. Las mismas fueron realizadas con el telescopio de 2.15m del Complejo Astronómico El Leoncito (CASLEO). El análisis de las velocidades radiales derivadas del material obtenido, confirma la realidad de Collinder 121, al menos desde el punto de vista cinemático. La velocidad radial baricentral (LSR) del cúmulo es de +17 ± 3 km.s-1. Esta velocidad coincide, dentro de los errores, con la velocidad radial (LSR) de la nebulosa anillo S308, la cual es de ~20 ± 10 km.s-1. Como S308 se encuentra físicamente asociada a la estrella Wolf-Rayet HD~50896, es muy probable que esta última sea un miembro de Collinder 121. Desde un punto de vista cinemático, la supergigante roja HD~50877 (K3Iab) también pertenecería a Collinder 121. Basándonos en la pertenencia de HD~50896 a Collinder 121, y en la interacción encontrada entre el viento de esta estrella y el medio interestelar circundante a la misma, se estima para este cúmulo una distancia del orden de 1 kpc.

Optimization of radially heterogeneous 1000-MW(e) LMFBR core configurations. Design and performance of reference cores. Research project 620-25

International Nuclear Information System (INIS)

Barthold, W.P.; Orechwa, Y.; Su, S.F.; Hutter, E.; Batch, R.V.; Beitel, J.C.; Turski, R.B.; Lam, P.S.K.

1979-11-01

A parameter study was conducted to determine the interrelated effects of: loosely of tightly coupled fuel regions separated by internal blanket assemblies, number of fuel regions, core height, number and arrangement of internal blanket subassemblies, number and size of fuel pins in a subassembly, etc. The effects of these parameters on sodium void reactivity, Doppler, incoherence, breeding gain, and thermohydraulics were of prime interest. Trends were established and ground work laid for optimization of a large, radially-heterogeneous, LMFBR core that will have low energetics in an HCDA and will have good thermal and breeding performance

A study on the nuclear characteristics of enriched gadolinia burnable absorber rods; the first year (2000) report

International Nuclear Information System (INIS)

Zee, Sung Quun; Lee, C.C.; Song, J. S.; Cho, B. O.; Joo, H. G.; Park, S. Y.; Kim, H. Y.; Cho, J. Y.; Kim, K. S.

2001-04-01

An analysis model using MICBURN-3/CASMO-3 is established for the enriched gadolinia burnable absorber rods. A homogenized cross section editing code, PROLOG, is modified so that it can handle such a fuel assembly that includes two different types of gadolinia rods. Study shows that Gd-155 and Gd-157 are almost same in suppressing the excess reactivity and it is recommended to enrich both odd number isotopes, Gd-155 and Gd-157. It is estimated that the cycle length increases by 2 days if enriched gadolinia rods are used in the commercial nuclear power plant such as YGN-3 of which the cycle length is assumed 2 years. For the advanced integral reactor SMART in which ultra long cycle length and soluble boron-free operation concept is applied, natural gadolinia burnable absorber rods fail to control the excess reactivity. On the other hand, enriched gadolinia rods are successful in controling the excess reactivity. To minimize power peakings, various placements of gadolinia rods are tested. Also initial reactivity holddown and gadolinia burnout time are parametrized with respect to the number of gadolinia rods and gadolinia weight fractions

Polarization switching and patterning in self-assembled peptide tubular structures

Science.gov (United States)

Bdikin, Igor; Bystrov, Vladimir; Delgadillo, Ivonne; Gracio, José; Kopyl, Svitlana; Wojtas, Maciej; Mishina, Elena; Sigov, Alexander; Kholkin, Andrei L.

2012-04-01

Self-assembled peptide nanotubes are unique nanoscale objects that have great potential for a multitude of applications, including biosensors, nanotemplates, tissue engineering, biosurfactants, etc. The discovery of strong piezoactivity and polar properties in aromatic dipeptides [A. Kholkin, N. Amdursky, I. Bdikin, E. Gazit, and G. Rosenman, ACS Nano 4, 610 (2010)] opened up a new perspective for their use as biocompatible nanoactuators, nanomotors, and molecular machines. Another, as yet unexplored functional property is the ability to switch polarization and create artificial polarization patterns useful in various electronic and optical applications. In this work, we demonstrate that diphenylalanine peptide nanotubes are indeed electrically switchable if annealed at a temperature of about 150 °C. The new orthorhombic antipolar structure that appears after annealing allows for the existence of a radial polarization component, which is directly probed by piezoresponse force microscopy (PFM) measurements. Observation of the relatively stable polarization patterns and hysteresis loops via PFM testifies to the local reorientation of molecular dipoles in the radial direction. The experimental results are complemented with rigorous molecular calculations and create a solid background of electric-field induced deformation of aromatic rings and corresponding polarization switching in this emergent material.

Blueprint for domestic uranium enrichment

International Nuclear Information System (INIS)

1981-01-01

The AEC advisory committee on domestic production of uranium enrichment has studied for more than a year how to achieve the domestic enrichment of uranium by the construction and operation of a commercial enriching plant using centrifugal separation method, and the report was submitted to the Atomic Energy Commission on August 18, 1980. Japan has depended wholly on overseas services for her uranium enrichment needs, but the development of domestic enrichment has been carried on in parallel. The AEC decided to construct a uranium enrichment pilot plant using centrifuges, and it has been forwarded as a national project. The plant is operated by the Power Reactor and Nuclear Fuel Development Corp. since 1979. The capacity of the plant will be raised to approximately 75 ton SWU a year. The centrifuges already operated have provided the first delivery of fuel of about 1 ton for the ATR ''Fugen''. The demand-supply balance of uranium enrichment service, the significance of the domestic enrichment of uranium, the evaluation of uranium enrichment technology, the target for domestic enrichment plan, the measures to promote domestic uranium enrichment, and the promotion of the construction of a demonstration plant are reported. (Kako, I.)

Radial supports of face motors with slack compensation

Energy Technology Data Exchange (ETDEWEB)

Kuznetsova, I I; Gelman, A B; Krekina, T V

1982-01-01

The design of a radial support of a face motor with slack compensation is described, and gives the results of field tests which confirm the performance capacity of the experimental support both from the viewpoint of durability, and in relation to preventing radial slack of the face motor shaft.

Queratotomía radial versus miniqueratotomía radial: Experiencia en el Hospital "Ramón Pando Ferrer" Radial keratotomy versus radial minikeratotomy: Experience in "Ramón Pando Ferrer" Hospital

Directory of Open Access Journals (Sweden)

José Edilberto Pacheco Serrano

2000-06-01

Full Text Available La miniqueratotomía radial se viene realizando desde 1995. Se plantea que incisiones más cortas tienen el mismo efecto y producen menos debilidad corneal, ya que disminuye la susceptibilidad a sufrir complicaciones graves provenientes de traumas de la vida cotidiana. Esta idea nos motivó a realizar un estudio para observar el comportamiento de incisiones más cortas en nuestro centro y, en caso de resultados positivos, implementar la técnica de manera que nuestros pacientes puedan beneficiarse de ella. Se comparan resultados de la aplicación de dos técnicas quirúrgicas refractivas para corrección de miopía entre leve y moderada. Se seleccionaron 38 pacientes entre 20 y 40 años de edad, con miopías entre -2 y -6 dioptrías y astigmatismo no mayor a -0,75 dioptrías. Se realizó queratotomía radial convencional en el ojo derecho y miniqueratotomía radial en el ojo izquierdo del mismo paciente. Las variaciones obtenidas en promedio fueron, en el ojo derecho: la esfera (en dioptrías D de -3,38 a -0,32, cilindro de -0,48 a -0,45 D, la queratometría de 44,75 a 41,21 D. En el ojo izquierdo: la esfera de -3,38 D a -0,44 D, cilindro de -0,44 D a -0,38 D, la queratometría de 44,83 a 41,80 D. Hubo una mejoría de la agudeza visual sin cristales de 0,61 en el ojo derecho y 0,59 en el ojo izquierdo. Las dos técnicas no mostraron diferencias estadísticamente significativas, con el beneficio de que la nueva técnica disminuye el riesgo de ruptura postraumática, según la bibliografía revisada, a causa de la menor injuria corneal.In this hospital, radial keratotomy is performed sice 1995. We propose that shorter incisions have some effect and cause less corneal weakness, since dicreases susceptibility to severe complications from traumata of daily life. This notion encouraged us to carry out a study to observe behaviour of shorter incisions in our service, and in the event of positive results, implementation of the technique so that our

Pseudarthrosis of radial shaft with dislocation of heads of radial and ulnar bones (case report

Directory of Open Access Journals (Sweden)

M. E. Puseva

2013-01-01

Full Text Available The authors presented a rare clinical case - the injury of forearm complicated by the formation of the pseudarthrosis of the radial shaft in combination with old dislocation of heads the radius and ulna. The differentiated approach to the choice of surgical tactics was proposed, which consists of several consistent stages: taking free autotransplant from the crest of iliac bone, resection of pseudarthrosis of radius with replacement of the bone defect by the graft for restoration of anatomic length, conducting combined strained osteosynthesis and elimination of dislocation of a head of radial and ulnar bones by transosseous osteosynthesis. The chosen treatment strategy allowed to restore the anatomy and function of the upper extremity.

Experimental critical parameters of enriched uranium solution in annular tank geometries

International Nuclear Information System (INIS)

Rothe, R.E.

1996-04-01

A total of 61 critical configurations are reported for experiments involving various combinations of annular tanks into which enriched uranium solution was pumped. These experiments were performed at two widely separated times in the 1980s under two programs at the Rocky Flats Plant's Critical Mass Laboratory. The uranyl nitrate solution contained about 370 g of uranium per liter, but this concentration varied a little over the duration of the studies. The uranium was enriched to about 93% [sup 235]U. All tanks were typical of sizes commonly found in nuclear production plants. They were about 2 m tall and ranged in diameter from 0.6 m to 1.5 m. Annular thicknesses and conditions of neutron reflection, moderation, and absorption were such that criticality would be achieved with these dimensions. Only 13 of the entire set of 74 experiments proved to be subcritical when tanks were completely filled with solution. Single tanks of several radial thicknesses were studied as well as small line arrays (1 x 2 and 1 x 3) of annular tanks. Many systems were reflected on four sides and the bottom by concrete, but none were reflected from above. Many experiments also contained materials within and outside the annular regions that contained strong neutron absorbers. One program had such a thick external moderator/absorber combination that no reflector was used at all

Experimental critical parameters of enriched uranium solution in annular tank geometries

Energy Technology Data Exchange (ETDEWEB)

Rothe, R.E.

1996-04-01

A total of 61 critical configurations are reported for experiments involving various combinations of annular tanks into which enriched uranium solution was pumped. These experiments were performed at two widely separated times in the 1980s under two programs at the Rocky Flats Plant`s Critical Mass Laboratory. The uranyl nitrate solution contained about 370 g of uranium per liter, but this concentration varied a little over the duration of the studies. The uranium was enriched to about 93% [sup 235]U. All tanks were typical of sizes commonly found in nuclear production plants. They were about 2 m tall and ranged in diameter from 0.6 m to 1.5 m. Annular thicknesses and conditions of neutron reflection, moderation, and absorption were such that criticality would be achieved with these dimensions. Only 13 of the entire set of 74 experiments proved to be subcritical when tanks were completely filled with solution. Single tanks of several radial thicknesses were studied as well as small line arrays (1 x 2 and 1 x 3) of annular tanks. Many systems were reflected on four sides and the bottom by concrete, but none were reflected from above. Many experiments also contained materials within and outside the annular regions that contained strong neutron absorbers. One program had such a thick external moderator/absorber combination that no reflector was used at all.

Comparison of the parameters of the IR-8 reactor with different fuel assembly designs with LEU fuel

International Nuclear Information System (INIS)

Vatulin, A.; Stetsky, Y.; Dobrikova, I.

1999-01-01

The estimation of neutron-physical, heat and hydraulic parameters of the IR-8 research reactor with low enriched uranium (LEU) fuel was performed. Two fuel assembly (FA) designs were reviewed: IRT-4M with the tubular type fuel elements and IRT-MR with the rod type fuel elements. UO 2 -Al dispersion 19.75% enrichment fuel is used in both cases. The results of the calculations were compared with main parameters of the reactor, using the current IRT-3M FA with 90% high enriched uranium (HEU) fuel. The results of these comparisons showed that during the LEU conversion of the reactor the cycle length, excess reactivity and peak power of the IRT-MR type FA are higher than for the IRT-3M type FA and IRT-4M type FA. (author)

Uranium enrichment. Enrichment processes

International Nuclear Information System (INIS)

Alexandre, M.; Quaegebeur, J.P.

2009-01-01

Despite the remarkable progresses made in the diversity and the efficiency of the different uranium enrichment processes, only two industrial processes remain today which satisfy all of enriched uranium needs: the gaseous diffusion and the centrifugation. This article describes both processes and some others still at the demonstration or at the laboratory stage of development: 1 - general considerations; 2 - gaseous diffusion: physical principles, implementation, utilisation in the world; 3 - centrifugation: principles, elementary separation factor, flows inside a centrifuge, modeling of separation efficiencies, mechanical design, types of industrial centrifuges, realisation of cascades, main characteristics of the centrifugation process; 4 - aerodynamic processes: vortex process, nozzle process; 5 - chemical exchange separation processes: Japanese ASAHI process, French CHEMEX process; 6 - laser-based processes: SILVA process, SILMO process; 7 - electromagnetic and ionic processes: mass spectrometer and calutron, ion cyclotron resonance, rotating plasmas; 8 - thermal diffusion; 9 - conclusion. (J.S.)

Promotion of uranium enrichment business

International Nuclear Information System (INIS)

Kurushima, Morihiro

1981-01-01

The Committee on Nuclear Power has studied on the basic nuclear power policy, establishing its five subcommittees, entrusted by the Ministry of Nternational Trade and Industry. The results of examination by the subcommittee on uranium enrichment business are given along with a report in this connection by the Committee. In order to establish the nuclear fuel cycle, the aspect of uranium enrichment is essential. The uranium enrichment by centrifugal process has proceeded steadily in Power Reactor and Nuclear Fuel Development Corporation. The following matters are described: the need for domestic uranium enrichment, the outlook for overseas enrichment services and the schedule for establishing domestic enrichment business, the current state of technology development, the position of the prototype enrichment plant, the course to be taken to establish enrichment business the main organization operating the prototype and commercial plants, the system of supplying centrifuges, the domestic conversion of natural uranium the subsidies for uranium enrichment business. (J.P.N.)

Effect of reactor size on the breeding economics of LMFBR blankets

International Nuclear Information System (INIS)

Tagishi, A.; Driscoll, M.J.

1975-02-01

The effect of reactor size on the neutronic and economic performance of LMFBR blankets driven by radially-power-flattened cores has been investigated using both simple models and state-of-the-art computer methods. Reactor power ratings in the range 250 to 3000 MW(e) were considered. Correlations for economic breakeven and optimum irradiation times and blanket thicknesses have been developed for batch-irradiated blankets. It is shown that a given distance from the core-blanket interface the fissile buildup rate per unit volume remains very nearly constant in the radial blanket as (radially-power-flattened, constant-height) core size increases. As a consequence, annual revenue per blanket assembly, and breakeven and optimum irradiation times and optimum blanket dimensions, are the same for all reactor sizes. It is also shown that the peripheral core fissile enrichment, hence neutron leakage spectra, of the (radially-power-flattened, constant-height) cores remains essentially constant as core size increases. Coupled with the preceding observations, this insures that radial blanket breeding performance in demonstration-size LMFBR units will be a good measure of that in much larger commercial LMFBR's

Calculation of Single Cell and Fuel Assembly IRIS Benchmarks Using WIMSD5B and GNOMER Codes

International Nuclear Information System (INIS)

Pevec, D.; Grgic, D.; Jecmenica, R.

2002-01-01

IRIS reactor (an acronym for International Reactor Innovative and Secure) is a modular, integral, light water cooled, small to medium power (100-335 MWe/module) reactor, which addresses the requirements defined by the United States Department of Energy for Generation IV nuclear energy systems, i.e., proliferation resistance, enhanced safety, improved economics, and waste reduction. An international consortium led by Westinghouse/BNFL was created for development of IRIS reactor; it includes universities, institutes, commercial companies, and utilities. Faculty of Electrical Engineering and Computing, University of Zagreb joined the consortium in year 2001, with the aim to take part in IRIS neutronics design and safety analyses of IRIS transients. A set of neutronic benchmarks for IRIS reactor was defined with the objective to compare results of all participants with exactly the same assumptions. In this paper a calculation of Benchmark 44 for IRIS reactor is described. Benchmark 44 is defined as a core depletion benchmark problem for specified IRIS reactor operating conditions (e.g., temperatures, moderator density) without feedback. Enriched boron, inhomogeneously distributed in axial direction, is used as an integral fuel burnable absorber (IFBA). The aim of this benchmark was to enable a more direct comparison of results of different code systems. Calculations of Benchmark 44 were performed using the modified CORD-2 code package. The CORD-2 code package consists of WIMSD and GNOMER codes. WIMSD is a well-known lattice spectrum calculation code. GNOMER solves the neutron diffusion equation in three-dimensional Cartesian geometry by the Green's function nodal method. The following parameters were obtained in Benchmark 44 analysis: effective multiplication factor as a function of burnup, nuclear peaking factor as a function of burnup, axial offset as a function of burnup, core-average axial power profile, core radial power profile, axial power profile for selected

Manufacturing of Precision Forgings by Radial Forging

International Nuclear Information System (INIS)

Wallner, S.; Harrer, O.; Buchmayr, B.; Hofer, F.

2011-01-01

Radial forging is a multi purpose incremental forging process using four tools on the same plane. It is widely used for the forming of tool steels, super alloys as well as titanium- and refractory metals. The range of application goes from reducing the diameters of shafts, tubes, stepped shafts and axels, as well as for creating internal profiles for tubes in Near-Net-Shape and Net-Shape quality. Based on actual development of a weight optimized transmission input shaft, the specific features of radial forging technology is demonstrated. Also a Finite Element Model for the simulation of the process is shown which leads to reduced pre-processing effort and reduced computing time compared to other published simulation methods for radial forging. The finite element model can be applied to quantify the effects of different forging strategies.

Reactive oxygen species promote heat shock protein 90-mediated HBV capsid assembly

International Nuclear Information System (INIS)

Kim, Yoon Sik; Seo, Hyun Wook; Jung, Guhung

2015-01-01

Hepatitis B virus (HBV) infection induces reactive oxygen species (ROS) production and has been associated with the development of hepatocellular carcinoma (HCC). ROS are also an important factor in HCC because the accumulated ROS leads to abnormal cell proliferation and chromosome mutation. In oxidative stress, heat shock protein 90 (Hsp90) and glutathione (GSH) function as part of the defense mechanism. Hsp90 prevents cellular component from oxidative stress, and GSH acts as antioxidants scavenging ROS in the cell. However, it is not known whether molecules regulated by oxidative stress are involved in HBV capsid assembly. Based on the previous study that Hsp90 facilitates HBV capsid assembly, which is an important step for the packing of viral particles, here, we show that ROS enrich Hsp90-driven HBV capsid formation. In cell-free system, HBV capsid assembly was facilitated by ROS with Hsp90, whereas it was decreased without Hsp90. In addition, GSH inhibited the function of Hsp90 to decrease HBV capsid assembly. Consistent with the result of cell-free system, ROS and buthionine sulfoximine (BS), an inhibitor of GSH synthesis, increased HBV capsid formation in HepG2.2.15 cells. Thus, our study uncovers the interplay between ROS and Hsp90 during HBV capsid assembly. - Highlights: • We examined H 2 O 2 and GSH modulate HBV capsid assembly. • H 2 O 2 facilitates HBV capsid assembly in the presence of Hsp90. • GSH inhibits function of Hsp90 in facilitating HBV capsid assembly. • H 2 O 2 and GSH induce conformation change of Hsp90

Reactive oxygen species promote heat shock protein 90-mediated HBV capsid assembly

Energy Technology Data Exchange (ETDEWEB)

Kim, Yoon Sik, E-mail: yumshak@naver.com; Seo, Hyun Wook, E-mail: suruk@naver.com; Jung, Guhung, E-mail: drjung@snu.ac.kr

2015-02-13

Hepatitis B virus (HBV) infection induces reactive oxygen species (ROS) production and has been associated with the development of hepatocellular carcinoma (HCC). ROS are also an important factor in HCC because the accumulated ROS leads to abnormal cell proliferation and chromosome mutation. In oxidative stress, heat shock protein 90 (Hsp90) and glutathione (GSH) function as part of the defense mechanism. Hsp90 prevents cellular component from oxidative stress, and GSH acts as antioxidants scavenging ROS in the cell. However, it is not known whether molecules regulated by oxidative stress are involved in HBV capsid assembly. Based on the previous study that Hsp90 facilitates HBV capsid assembly, which is an important step for the packing of viral particles, here, we show that ROS enrich Hsp90-driven HBV capsid formation. In cell-free system, HBV capsid assembly was facilitated by ROS with Hsp90, whereas it was decreased without Hsp90. In addition, GSH inhibited the function of Hsp90 to decrease HBV capsid assembly. Consistent with the result of cell-free system, ROS and buthionine sulfoximine (BS), an inhibitor of GSH synthesis, increased HBV capsid formation in HepG2.2.15 cells. Thus, our study uncovers the interplay between ROS and Hsp90 during HBV capsid assembly. - Highlights: • We examined H{sub 2}O{sub 2} and GSH modulate HBV capsid assembly. • H{sub 2}O{sub 2} facilitates HBV capsid assembly in the presence of Hsp90. • GSH inhibits function of Hsp90 in facilitating HBV capsid assembly. • H{sub 2}O{sub 2} and GSH induce conformation change of Hsp90.

Inspection experience with RA-3 spent nuclear fuel assemblies at CNEA's central storage facility

International Nuclear Information System (INIS)

Novara, Oscar; LaFuente, Jose; Large, Steve; Andes, Trent; Messick, Charles

2000-01-01

Aluminum-based spent nuclear fuel from Argentina's RA-3 research reactor is to be shipped to the Savannah River Site near Aiken, South Carolina, USA. The spent nuclear fuel contains highly enriched uranium of U.S. origin and is being returned under the US Department of Energy's Foreign Research Reactor/Domestic Research Reactor (FRR/DRR) Receipt Program. An intensive inspection of 207 stored fuel assemblies was conducted to assess shipping cask containment limitations and assembly handling considerations. The inspection was performed with video equipment designed for remote operation, high portability, easy setup and usage. Fuel assemblies were raised from their vertical storage tubes, inspected by remote video, and then returned to their original storage tube or transferred to an alternate location. The inspections were made with three simultaneous video systems, each with dedicated viewing, digital recording, and tele-operated control from a shielded location. All 207 fuel assemblies were safely and successfully inspected in fifteen working days. Total dose to personnel was about one-half of anticipated dose. (author)

Radially truncated galactic discs

NARCIS (Netherlands)

Grijs, R. de; Kregel, M.; Wesson, K H

2000-01-01

Abstract: We present the first results of a systematic analysis of radially truncatedexponential discs for four galaxies of a sample of disc-dominated edge-onspiral galaxies. Edge-on galaxies are very useful for the study of truncatedgalactic discs, since we can follow their light distributions out

From high enriched to low enriched uranium fuel in research reactors

Energy Technology Data Exchange (ETDEWEB)

Van Den Berghe, S.; Leenaers, A.; Koonen, E.; Moons, F.; Sannen, L. [Nuclear Materials Science Institute, SCK.CEN, Boeretang 200, B-2400 Mol (Belgium)

2010-07-01

Since the 1970's, global efforts have been going on to replace the high-enriched (>90% {sup 235}U), low-density UAlx research reactor fuel with high-density, low enriched (<20% {sup 235}U) replacements. This search is driven by the attempt to reduce the civil use of high-enriched material because of proliferation risks and terrorist threats. American initiatives, such as the Global Threat Reduction Initiative (GTRI) and the Reduced Enrichment for Research and Test Reactors (RERTR) program have triggered the development of reliable low-enriched fuel types for these reactors, which can replace the high enriched ones without loss of performance. Most success has presently been obtained with U{sub 3}Si{sub 2} dispersion fuel, which is currently used in many research reactors in the world. However, efforts to search for a replacement with even higher density, which will also allow the conversion of some high flux research reactors that currently cannot change to U{sub 3}Si{sub 2} (eg. BR2 in Belgium), have continued and are for the moment mainly directed towards the U(Mo) alloy fuel (7-10 w% Mo). This paper provides an overview of the past efforts and presents the current status of the U(Mo) development. (authors)

From high enriched to low enriched uranium fuel in research reactors

International Nuclear Information System (INIS)

Van Den Berghe, S.; Leenaers, A.; Koonen, E.; Moons, F.; Sannen, L.

2010-01-01

Since the 1970's, global efforts have been going on to replace the high-enriched (>90% 235 U), low-density UAlx research reactor fuel with high-density, low enriched ( 235 U) replacements. This search is driven by the attempt to reduce the civil use of high-enriched material because of proliferation risks and terrorist threats. American initiatives, such as the Global Threat Reduction Initiative (GTRI) and the Reduced Enrichment for Research and Test Reactors (RERTR) program have triggered the development of reliable low-enriched fuel types for these reactors, which can replace the high enriched ones without loss of performance. Most success has presently been obtained with U 3 Si 2 dispersion fuel, which is currently used in many research reactors in the world. However, efforts to search for a replacement with even higher density, which will also allow the conversion of some high flux research reactors that currently cannot change to U 3 Si 2 (eg. BR2 in Belgium), have continued and are for the moment mainly directed towards the U(Mo) alloy fuel (7-10 w% Mo). This paper provides an overview of the past efforts and presents the current status of the U(Mo) development. (authors)

Uranium enrichment by gas centrifuge

International Nuclear Information System (INIS)

Heriot, I.D.

1988-01-01

After recalling the physical principles and the techniques of centrifuge enrichment the report describes the centrifuge enrichment programmes of the various countries concerned and compares this technology with other enrichment technologies like gaseous diffusion, laser, aerodynamic devices and chemical processes. The centrifuge enrichment process is said to be able to replace with advantage the existing enrichment facilities in the short and medium term. Future prospects of the process are also described, like recycled uranium enrichment and economic improvements; research and development needs to achieve the economic prospects are also indicated. Finally the report takes note of the positive aspect of centrifuge enrichment as far as safeguards and nuclear safety are concerned. 27 figs, 113 refs

Enriching Genomic Resources and Marker Development from Transcript Sequences of Jatropha curcas for Microgravity Studies

Science.gov (United States)

Tian, Wenlan; Paudel, Dev

2017-01-01

Jatropha (Jatropha curcas L.) is an economically important species with a great potential for biodiesel production. To enrich the jatropha genomic databases and resources for microgravity studies, we sequenced and annotated the transcriptome of jatropha and developed SSR and SNP markers from the transcriptome sequences. In total 1,714,433 raw reads with an average length of 441.2 nucleotides were generated. De novo assembling and clustering resulted in 115,611 uniquely assembled sequences (UASs) including 21,418 full-length cDNAs and 23,264 new jatropha transcript sequences. The whole set of UASs were fully annotated, out of which 59,903 (51.81%) were assigned with gene ontology (GO) term, 12,584 (10.88%) had orthologs in Eukaryotic Orthologous Groups (KOG), and 8,822 (7.63%) were mapped to 317 pathways in six different categories in Kyoto Encyclopedia of Genes and Genome (KEGG) database, and it contained 3,588 putative transcription factors. From the UASs, 9,798 SSRs were discovered with AG/CT as the most frequent (45.8%) SSR motif type. Further 38,693 SNPs were detected and 7,584 remained after filtering. This UAS set has enriched the current jatropha genomic databases and provided a large number of genetic markers, which can facilitate jatropha genetic improvement and many other genetic and biological studies. PMID:28154822

Derived enriched uranium market

International Nuclear Information System (INIS)

Rutkowski, E.

1996-01-01

The potential impact on the uranium market of highly enriched uranium from nuclear weapons dismantling in the Russian Federation and the USA is analyzed. Uranium supply, conversion, and enrichment factors are outlined for each country; inventories are also listed. The enrichment component and conversion components are expected to cause little disruption to uranium markets. The uranium component of Russian derived enriched uranium hexafluoride is unresolved; US legislation places constraints on its introduction into the US market

Radial velocity observations of VB10

Science.gov (United States)

Deshpande, R.; Martin, E.; Zapatero Osorio, M. R.; Del Burgo, C.; Rodler, F.; Montgomery, M. M.

2011-07-01

VB 10 is the smallest star known to harbor a planet according to the recent astrometric study of Pravdo & Shaklan [1]. Here we present near-infrared (J-band) radial velocity of VB 10 performed from high resolution (R~20,000) spectroscopy (NIRSPEC/KECK II). Our results [2] suggest radial velocity variability with amplitude of ~1 km/s, a result that is consistent with the presence of a massive planet companion around VB10 as found via long-term astrometric monitoring of the star by Pravdo & Shaklan. Employing an entirely different technique we verify the results of Pravdo & Shaklan.

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.

Radial electron beam laser excitation: the REBLE report

International Nuclear Information System (INIS)

Ramirez, J.J.; Prestwich, K.R.

1978-10-01

The results of an investigation of techniques to generate high-power radially converging electron beams and the application of these beams to gas lasers is discussed. The design and performance of the REBLE accelerator that was developed for this program is presented. Reliable operation of the radial diode has been obtained at levels up to 1 MV, 200 kA, and 20 ns. It has been demonstrated that the anode current density can be made uniform to better than 15% over 1000 cm 2 areas with 100 to 250 A/cm 2 intensities. The measured total and spatially resolved energy deposition of this radial electron beam in various gases is compared with Monte Carlo calculations. In most cases, these codes give an accurate description of the beam transport and energy deposition. With the electron beam pumping xenon gas, the amplitude of xenon excimer radiation (1720 A 0 ) was radially uniform to within the experimental uncertainty. The efficiency of converting deposited electron beam energy to xenon excimer radiation was 20%

DESIGN STUDY FOR A LOW-ENRICHED URANIUM CORE FOR THE HIGH FLUX ISOTOPE REACTOR, ANNUAL REPORT FOR FY 2010

Energy Technology Data Exchange (ETDEWEB)

Cook, David Howard [ORNL; Freels, James D [ORNL; Ilas, Germina [ORNL; Jolly, Brian C [ORNL; Miller, James Henry [ORNL; Primm, Trent [ORNL; Renfro, David G [ORNL; Sease, John D [ORNL; Pinkston, Daniel [ORNL

2011-02-01

This report documents progress made during FY 2010 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current level. Studies are reported of support to a thermal hydraulic test loop design, the implementation of finite element, thermal hydraulic analysis capability, and infrastructure tasks at HFIR to upgrade the facility for operation at 100 MW. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. Continuing development in the definition of the fuel fabrication process is described.

Radial scar/complex sclerosing lesion of the breast--value of ultrasound.

Science.gov (United States)

Grunwald, S; Heyer, H; Kühl, A; Schwesinger, G; Schimming, A; Köhler, G; Ohlinger, R

2007-04-01

Although benign, radial scar/complex sclerosing adenosis is a lesion which histopathologically resembles tubular carcinoma. On physical examination, it is difficult to distinguish radial scar from a malignant tumour. Mammography cannot differentiate radial scar from malignancy. This clinical study aims to delineate the role of preoperative ultrasonography with emphasis on the question whether ultrasonography could lower the number of false-positive readings and therefore the number of open biopsies required. In this examination, we present the clinical, mammographic, ultrasonographic, and histopathological features of 6 cases of radial scars. Although most authors describe radial scars as non-palpable, 2 of 6 lesions were indeed palpable. On mammograms, radial scars have a spiculated appearance, a feature observed in all of our cases. Numerous ultrasonographic characteristics are listed in the literature, but ultrasonography is not reported to have clear-cut advantages. Although this study did not elucidate any unique ultrasonographic features to characterise these lesions, the analysis of all ultrasonographic results made us recognise a set of "nearly specific ultrasonographic features" of radial scars. Current B-mode imaging does not appear to lead to the desirable reduction of the rate of unnecessary open biopsies.

Radial Structure Scaffolds Convolution Patterns of Developing Cerebral Cortex

Directory of Open Access Journals (Sweden)

Mir Jalil Razavi

2017-08-01

Full Text Available Commonly-preserved radial convolution is a prominent characteristic of the mammalian cerebral cortex. Endeavors from multiple disciplines have been devoted for decades to explore the causes for this enigmatic structure. However, the underlying mechanisms that lead to consistent cortical convolution patterns still remain poorly understood. In this work, inspired by prior studies, we propose and evaluate a plausible theory that radial convolution during the early development of the brain is sculptured by radial structures consisting of radial glial cells (RGCs and maturing axons. Specifically, the regionally heterogeneous development and distribution of RGCs controlled by Trnp1 regulate the convex and concave convolution patterns (gyri and sulci in the radial direction, while the interplay of RGCs' effects on convolution and axons regulates the convex (gyral convolution patterns. This theory is assessed by observations and measurements in literature from multiple disciplines such as neurobiology, genetics, biomechanics, etc., at multiple scales to date. Particularly, this theory is further validated by multimodal imaging data analysis and computational simulations in this study. We offer a versatile and descriptive study model that can provide reasonable explanations of observations, experiments, and simulations of the characteristic mammalian cortical folding.

Californium interrogation prompt neutron (CIPN) instrument for non-destructive assay of spent nuclear fuel—Design concept and experimental demonstration

International Nuclear Information System (INIS)

Henzlova, D.; Menlove, H.O.; Rael, C.D.; Trellue, H.R.; Tobin, S.J.; Park, Se-Hwan; Oh, Jong-Myeong; Lee, Seung-Kyu; Ahn, Seong-Kyu; Kwon, In-Chan; Kim, Ho-Dong

2016-01-01

This paper presents results of the first experimental demonstration of the Californium Interrogation Prompt Neutron (CIPN) instrument developed within a multi-year effort launched by the Next Generation Safeguards Initiative Spent Fuel Project of the United States Department of Energy. The goals of this project focused on developing viable non-destructive assay techniques with capabilities to improve an independent verification of spent fuel assembly characteristics. For this purpose, the CIPN instrument combines active and passive neutron interrogation, along with passive gamma-ray measurements, to provide three independent observables. This paper describes the initial feasibility demonstration of the CIPN instrument, which involved measurements of four pressurized-water-reactor spent fuel assemblies with different levels of burnup and two initial enrichments. The measurements were performed at the Post-Irradiation Examination Facility at the Korea Atomic Energy Institute in the Republic of Korea. The key aim of the demonstration was to evaluate CIPN instrument performance under realistic deployment conditions, with the focus on a detailed assessment of systematic uncertainties that are best evaluated experimentally. The measurements revealed good positioning reproducibility, as well as a high degree of insensitivity of the CIPN instrument's response to irregularities in a radial burnup profile. Systematic uncertainty of individual CIPN instrument signals due to assembly rotation was found to be <4.5%, even for assemblies with fairly extreme gradients in the radial burnup profile. These features suggest that the CIPN instrument is capable of providing a good representation of assembly average characteristics, independent of assembly orientation in the instrument.

Californium interrogation prompt neutron (CIPN) instrument for non-destructive assay of spent nuclear fuel—Design concept and experimental demonstration

Energy Technology Data Exchange (ETDEWEB)

Henzlova, D., E-mail: henzlova@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Menlove, H.O.; Rael, C.D.; Trellue, H.R.; Tobin, S.J. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Park, Se-Hwan; Oh, Jong-Myeong; Lee, Seung-Kyu; Ahn, Seong-Kyu; Kwon, In-Chan; Kim, Ho-Dong [Korea Atomic Energy Research Institute, Daejeong (Korea, Republic of)

2016-01-11

This paper presents results of the first experimental demonstration of the Californium Interrogation Prompt Neutron (CIPN) instrument developed within a multi-year effort launched by the Next Generation Safeguards Initiative Spent Fuel Project of the United States Department of Energy. The goals of this project focused on developing viable non-destructive assay techniques with capabilities to improve an independent verification of spent fuel assembly characteristics. For this purpose, the CIPN instrument combines active and passive neutron interrogation, along with passive gamma-ray measurements, to provide three independent observables. This paper describes the initial feasibility demonstration of the CIPN instrument, which involved measurements of four pressurized-water-reactor spent fuel assemblies with different levels of burnup and two initial enrichments. The measurements were performed at the Post-Irradiation Examination Facility at the Korea Atomic Energy Institute in the Republic of Korea. The key aim of the demonstration was to evaluate CIPN instrument performance under realistic deployment conditions, with the focus on a detailed assessment of systematic uncertainties that are best evaluated experimentally. The measurements revealed good positioning reproducibility, as well as a high degree of insensitivity of the CIPN instrument's response to irregularities in a radial burnup profile. Systematic uncertainty of individual CIPN instrument signals due to assembly rotation was found to be <4.5%, even for assemblies with fairly extreme gradients in the radial burnup profile. These features suggest that the CIPN instrument is capable of providing a good representation of assembly average characteristics, independent of assembly orientation in the instrument.

Low-enriched uranium high-density target project. Compendium report

Energy Technology Data Exchange (ETDEWEB)

Vandegrift, George; Brown, M. Alex; Jerden, James L.; Gelis, Artem V.; Stepinski, Dominique C.; Wiedmeyer, Stanley; Youker, Amanda; Hebden, Andrew; Solbrekken, G; Allen, C; Robertson., D; El-Gizawy, Sherif; Govindarajan, Srisharan; Hoyer, Annemarie; Makarewicz, Philip; Harris, Jacob; Graybill, Brian; Gunn, Andy; Berlin, James; Bryan, Chris; Sherman, Steven; Hobbs, Randy; Griffin, F. P.; Chandler, David; Hurt, C. J.; Williams, Paul; Creasy, John; Tjader, Barak; McFall, Danielle; Longmire, Hollie

2016-09-01

At present, most ⁹⁹Mo is produced in research, test, or isotope production reactors by irradiation of highly enriched uranium targets. To achieve the denser form of uranium needed for switching from high to low enriched uranium (LEU), targets in the form of a metal foil (~125-150 µm thick) are being developed. The LEU High Density Target Project successfully demonstrated several iterations of an LEU-fission-based Mo-99 technology that has the potential to provide the world’s supply of Mo-99, should major producers choose to utilize the technology. Over 50 annular high density targets have been successfully tested, and the assembly and disassembly of targets have been improved and optimized. Two target front-end processes (acidic and electrochemical) have been scaled up and demonstrated to allow for the high-density target technology to mate up to the existing producer technology for target processing. In the event that a new target processing line is started, the chemical processing of the targets is greatly simplified. Extensive modeling and safety analysis has been conducted, and the target has been qualified to be inserted into the High Flux Isotope Reactor, which is considered above and beyond the requirements for the typical use of this target due to high fluence and irradiation duration.

300 MWe Burner Core Design with two Enrichment Zoning

International Nuclear Information System (INIS)

Song, Hoon; Kim, Sang Ji; Kim, Yeong Il

2008-01-01

KAERI has been developing the KALIMER-600 core design with a breakeven fissile conversion ratio. The core is loaded with a ternary metallic fuel (TRU-U-10Zr), and the breakeven characteristics are achieved without any blanket assembly. As an alternative plan, a KALIMER-600 burner core design has been also performed. In the early stage of the development of a fast reactor, the main purpose is an economical use of a uranium resource but nowadays in addition to the maximum utilization of a uranium resource, the burning of a high level radioactive waste is taken as an additional interest for the harmony of the environment. In way of constructing the commercial size reactor which has the power level ranging from 800 MWe to 1600 MWe, the demonstration reactor which has the power level ranging from 200 MWe to 600 MWe was usually constructed for the midterm stage to commercial size reactor. In this paper, a 300 MWe burner core design was performed with purpose of demonstration reactor for KALIMER-600 burner of 600 MWe. As a means to flatten the power distribution, instead of a single fuel enrichment scheme adapted in design of KALIMER-600 burner, the 2 enrichment zoning approach was adapted

Application of genetic algorithms and CASMO to fuel optimization of BWRs

International Nuclear Information System (INIS)

Carmona H, R.; Martin del Campo M, C.; Oropeza C, I.P.

2008-01-01

It was developed a system for the optimization of the radial distribution of enrichment in a fuel cell of a boiling water reactor based on genetic algorithms (GA's). The objective function includes four parameters: Average of the cell enrichment, average of gadolinium concentration of the cell, radial peak power factor and multiplication k-infinite factor. In order to be able to calculate the parameters that take part in the objective function, the process of evaluation of GA's was tied to the code CASMO-4, which is a code of transport in neutronic simulation groups of fuel assemblies that have been validated and it is used thoroughly for the calculation of nuclear data banks for boiling water reactors. A good radial distribution of fuel rods looks for, with different enrichment of U 2 35 and contents of consumable poison in gadolinium form. For it is necessary to define the representation of the solution, the objective function and the implementation of the specific optimization process to the solution problem. The optimization process was codified in language C in the operating system LINUX. It was automated the creation of the entrances to the simulator, the execution of simulator CASMO-4 and the obtaining of the parameters that take part in the objective function from the exit of the simulator. It was applied to the fuel cell design of lOxlO that can be used in the fuel designs which are used at the moment in the nuclear power plant of Laguna Verde. They were considered 10 different fuel compositions from which four contain gadolinium. Three heuristic rules were applied: the peripheral positions of the fuel cell cannot contain burning poison, are placed the compositions with the smallest enrichment in the cell corners and, it is fixed the placement of the water rods. Nevertheless, the placement of the rods with gadolinium cell inside left free. Designs were obtained that complete with the wanted reactivity and the radial peak power factor. The best in them has

Radiometric enrichment of nonradioactive ores

International Nuclear Information System (INIS)

Mokrousov, V.A.; Lileev, V.A.

1979-01-01

Considered are the methods of mineral enrichment based on the use of the radioation of various types. The physical essence of enrichment processes is presented, their classification is given. Described are the ore properties influencing the efficiency of radiometric enrichment, methods of the properties study and estimation of ore enrichment. New possibilities opened by radiometric enrichment in the technology of primary processing of mineral raw materials are elucidated. A considerable attention is paid to the main and auxiliary equipment for radiometric enrichment. The foundations of the safety engineering are presented in a brief form. Presented are also results of investigations and practical works in the field of enrichment of ores of non-ferrous, ferrous and non-metallic minerals with the help of radiometric methods

Nuclear fuel assembly grid sleeve/guide thimble bulge orientation gage and inspection method

International Nuclear Information System (INIS)

Widener, W.H.

1988-01-01

This patent describes a method of inspecting a fuel assembly to determine the orientation of externally-projecting mated bulges connecting a grid sleeve to a guide thimble of the assembly, the method comprising the steps of: (a) inserting a radially-expandable tubular member within the guide thimble, the tubular member having externally-projecting embossments thereon spaced circumferentially from one another about the tubular member, the embossments being the same in number as the bulges of the guide thimble and configured to fit therewithin; (b) axially moving an elongated expansion member, which extends through and rotatably mounts the tubular member, relative to the tubular member from a first position in which the expansion member permits inward contraction of the tubular member and displacement of embossments thereon away from the interior of the guide thimble bulges for removing the embossments from registry therewith and a second position in which the expansion member produces radial expansion of the tubular member and displacement of the embossments thereon toward the interior of the guide thimble bulges for placing the embossments in registry therewith; (c) rotating the tubular member relative to the expansion member so as to bring the embossments on the tubular member into alignment with the guide thimble bulges as the embossments on the tubular member are being displaced toward and into registry with the interior of the bulges; and (d) responsive to rotation of the tubular member away from a reference position, providing an indication of the orientation of the guide thimble bulges relative to a reference point upon displacement of the embossments into registry therewith

77 FR 14838 - General Electric-Hitachi Global Laser Enrichment LLC, Commercial Laser-Based Uranium Enrichment...

Science.gov (United States)

2012-03-13

... Laser Enrichment LLC, Commercial Laser-Based Uranium Enrichment Facility, Wilmington, North Carolina... a license to General Electric-Hitachi Global Laser Enrichment LLC (GLE or the applicant) to authorize construction of a laser-based uranium enrichment facility and possession and use of byproduct...

Low-Enriched Uranium Fuel Conversion Activities for the High Flux Isotope Reactor, Annual Report for FY 2011

Energy Technology Data Exchange (ETDEWEB)

Renfro, David G [ORNL; Cook, David Howard [ORNL; Freels, James D [ORNL; Griffin, Frederick P [ORNL; Ilas, Germina [ORNL; Sease, John D [ORNL; Chandler, David [ORNL

2012-03-01

This report describes progress made during FY11 in ORNL activities to support converting the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum (UMo) alloy. With both radial and axial contouring of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current levels achieved with HEU fuel. Studies are continuing to demonstrate that the fuel thermal safety margins can be preserved following conversion. Studies are also continuing to update other aspects of the reactor steady state operation and accident response for the effects of fuel conversion. Technical input has been provided to Oregon State University in support of their hydraulic testing program. The HFIR conversion schedule was revised and provided to the GTRI program. In addition to HFIR conversion activities, technical support was provided directly to the Fuel Fabrication Capability program manager.

Low-Enriched Uranium Fuel Conversion Activities for the High Flux Isotope Reactor, Annual Report for FY 2011

International Nuclear Information System (INIS)

Renfro, David G.; Cook, David Howard; Freels, James D.; Griffin, Frederick P.; Ilas, Germina; Sease, John D.; Chandler, David

2012-01-01

This report describes progress made during FY11 in ORNL activities to support converting the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum (UMo) alloy. With both radial and axial contouring of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current levels achieved with HEU fuel. Studies are continuing to demonstrate that the fuel thermal safety margins can be preserved following conversion. Studies are also continuing to update other aspects of the reactor steady state operation and accident response for the effects of fuel conversion. Technical input has been provided to Oregon State University in support of their hydraulic testing program. The HFIR conversion schedule was revised and provided to the GTRI program. In addition to HFIR conversion activities, technical support was provided directly to the Fuel Fabrication Capability program manager.

Leading-process actomyosin coordinates organelle positioning and adhesion receptor dynamics in radially migrating cerebellar granule neurons.

Science.gov (United States)

Trivedi, Niraj; Ramahi, Joseph S; Karakaya, Mahmut; Howell, Danielle; Kerekes, Ryan A; Solecki, David J

2014-12-02

During brain development, neurons migrate from germinal zones to their final positions to assemble neural circuits. A unique saltatory cadence involving cyclical organelle movement (e.g., centrosome motility) and leading-process actomyosin enrichment prior to nucleokinesis organizes neuronal migration. While functional evidence suggests that leading-process actomyosin is essential for centrosome motility, the role of the actin-enriched leading process in globally organizing organelle transport or traction forces remains unexplored. We show that myosin ii motors and F-actin dynamics are required for Golgi apparatus positioning before nucleokinesis in cerebellar granule neurons (CGNs) migrating along glial fibers. Moreover, we show that primary cilia are motile organelles, localized to the leading-process F-actin-rich domain and immobilized by pharmacological inhibition of myosin ii and F-actin dynamics. Finally, leading process adhesion dynamics are dependent on myosin ii and F-actin. We propose that actomyosin coordinates the overall polarity of migrating CGNs by controlling asymmetric organelle positioning and cell-cell contacts as these cells move along their glial guides.

SIMMER-III parametric studies of fuel-steel mixing and radial mesh effects on power excursion in ESFR ULOF transients - 15033

International Nuclear Information System (INIS)

Chen, X.N.; Rineiski, A.; Gabrielli, F.; Andriolo, L.; Li, R.; Maschek, W.

2015-01-01

This paper deals with SIMMER-III once-through simulations of the first power excursion initiated by an unprotected loss of flow (ULOF) in the Working Horse design of the European Sodium Cooled Fast Reactor (ESFR). Since the sodium void effect is strictly positive in this core and dominant in the transient, a power excursion is initiated by sodium boiling in the ULOF case. Two major effects, namely (1) reactivity effects due to fuel-steel mixing after melting and (2) the radial mesh size, which were not considered initially in SIMMER simulations for ESFR, are studied. The first effect concerns the reactivity difference between the heterogeneous fuel/clad/wrapper configuration and the homogeneous mixture of steel and fuel. The full core homogenization (due to melting) effect is ∼ 2 dollars, though a smaller effect takes place in case of partial core melting. The second effect is due to the SIMMER sub-assembly coarse mesh treatment, where a simultaneous sodium boiling onset in all sub-assemblies belonging to one ring leads to an overestimated reactivity ramp. For investigating the influence of fuel/steel mixing effects, a lumped 'homogenization' reactivity feedback has been introduced, being proportional to the molten steel mass. For improving the coarse mesh treatment, we employ finer radial meshes to take the subchannel effects into account, where the side and interior channels have different coolant velocities and temperatures. The simulation results show that these two effects have significant impacts on the first power excursion after the sodium boiling: both effects delay the power excursion and significantly reduce the height of the power peaks in case of a ULOF

Analysis of subcritical control rod worth measurements in assembly BZB/3

International Nuclear Information System (INIS)

Giese, H.

1981-07-01

A series of subcritical absorber array measurements was performed in version three of the BIZET assembly BZB in order to check the ability of standard reactor computational codes used by the BIZET participants in predicting control rod worths in large fast reactors. Assembly BZB/3 was a two-zone core with a diameter of about 2.5 m and a core height of 0.89 m, fuelled with plutonium. Fifteen control rod positions and twelve secondary shutdown rod positions were simulated in the core. The measurements comprised the insertion of single absorbers as well as various groups of absorbers and were based on the modified source multiplication method. The KfK analysis was confined to the calculation of eigenvalues for different absorber arrays, also with a view to a comparison with the results of a former BZA evaluation with calculation-to-experiment values of up to C/E ∼ 1.10. The C/E-values found for BZB/3 ranged from 1.02 to 1.10 and did not show a systematic variation at different radial positions or different degrees of absorber asymmetry

Stereochemistry in subcomponent self-assembly.

Science.gov (United States)

Castilla, Ana M; Ramsay, William J; Nitschke, Jonathan R

2014-07-15

CONSPECTUS: As Pasteur noted more than 150 years ago, asymmetry exists in matter at all organization levels. Biopolymers such as proteins or DNA adopt one-handed conformations, as a result of the chirality of their constituent building blocks. Even at the level of elementary particles, asymmetry exists due to parity violation in the weak nuclear force. While the origin of homochirality in living systems remains obscure, as does the possibility of its connection with broken symmetries at larger or smaller length scales, its centrality to biomolecular structure is clear: the single-handed forms of bio(macro)molecules interlock in ways that depend upon their handednesses. Dynamic artificial systems, such as helical polymers and other supramolecular structures, have provided a means to study the mechanisms of transmission and amplification of stereochemical information, which are key processes to understand in the context of the origins and functions of biological homochirality. Control over stereochemical information transfer in self-assembled systems will also be crucial for the development of new applications in chiral recognition and separation, asymmetric catalysis, and molecular devices. In this Account, we explore different aspects of stereochemistry encountered during the use of subcomponent self-assembly, whereby complex structures are prepared through the simultaneous formation of dynamic coordinative (N → metal) and covalent (N═C) bonds. This technique provides a useful method to study stereochemical information transfer processes within metal-organic assemblies, which may contain different combinations of fixed (carbon) and labile (metal) stereocenters. We start by discussing how simple subcomponents with fixed stereogenic centers can be incorporated in the organic ligands of mononuclear coordination complexes and communicate stereochemical information to the metal center, resulting in diastereomeric enrichment. Enantiopure subcomponents were then

Application of SN and Monte Carlo codes to the SHEBA critical assemblies

International Nuclear Information System (INIS)

O'Dell, R.D.

1993-01-01

The Solution High-Energy Burst Assembly (SHEBA) at Los Alamos is a low-enriched (4.95 wt. %) aqueous uranyl fluoride solution critical assembly. There are two SHEBA configurations, both consisting of right circular cylinders with a central control rod. The first configuration, hereafter called the old SHEBA, had a fuel solution diameter of 54.6 cm and a measured critical solution height of 36.5 cm. An improved modification, hereafter called the new SHEBA, has a fuel solution diameter of 48.9 cm but since it is not yet operational, the critical solution height has not yet been measured. In this presentation the application of the discrete-ordinates (S N ) code TWODANT using Hansen-Roach cross sections and the MCNP Monte Carlo code using continuous-energy cross sections for calculating the critical solution heights for both the old and new SHEBA assemblies is described. The code's predictions are compared and it is shown that a single calculation with a standard computer code may yield misleading results, especially when using a Monte Carlo code

Numerical Simulation of Hydraulic Fracture Propagation Guided by Single Radial Boreholes

Directory of Open Access Journals (Sweden)

Tiankui Guo

2017-10-01

Full Text Available Conventional hydraulic fracturing is not effective in target oil development zones with available wellbores located in the azimuth of the non-maximum horizontal in-situ stress. To some extent, we think that the radial hydraulic jet drilling has the function of guiding hydraulic fracture propagation direction and promoting deep penetration, but this notion currently lacks an effective theoretical support for fracture propagation. In order to verify the technology, a 3D extended finite element numerical model of hydraulic fracturing promoted by the single radial borehole was established, and the influences of nine factors on propagation of hydraulic fracture guided by the single radial borehole were comprehensively analyzed. Moreover, the term ‘Guidance factor (Gf’ was introduced for the first time to effectively quantify the radial borehole guidance. The guidance of nine factors was evaluated through gray correlation analysis. The experimental results were consistent with the numerical simulation results to a certain extent. The study provides theoretical evidence for the artificial control technology of directional propagation of hydraulic fracture promoted by the single radial borehole, and it predicts the guidance effect of a single radial borehole on hydraulic fracture to a certain extent, which is helpful for planning well-completion and fracturing operation parameters in radial borehole-promoted hydraulic fracturing technology.

Radial collective flow in heavy-ion collisions at intermediate energies

International Nuclear Information System (INIS)

Borderie, B.

1996-11-01

The production of radial collective flow is associated with collisions leading to sources which undergo multifragmentation/explosion processes. After a theoretical survey of possible causes of production of radial flow, methods used to derive experimental values are discussed. Finally, a large set of data is presented which can be used to study and disentangle the different effects leading to radial collective flow. The dominant role of compression in the lower energy domain is emphasized. (author)

Rotary and radial forcing effects on center-of-mass locomotion dynamics.

Science.gov (United States)

Shen, Z H; Larson, P L; Seipel, J E

2014-09-01

Rotary and radial forcing are two common actuation methods for legged robots. However, these two orthogonal methods of center-of-mass (CoM) forcing have not been compared as potentially alternative strategies of actuation. In this paper, we compare the CoM stability and energetics of running with rotary and radial actuation through the simulation of two models: the rotary-forced spring-loaded inverted pendulum (rotary-forced-SLIP), and the radially-forced-SLIP. We model both radial and rotary actuation in the simplest way, applying them as a constant force during the stance portion of the gait. A simple application of constant rotary forcing throughout stance is capable of producing fully-asymptotically stable motion; however, a similarly constant application of radial forcing throughout the stance is not capable of producing stable solutions. We then allow both the applied rotary and radial forcing functions to turn on or off based on the occurrence of the mid-stance event, which breaks the symmetry of actuation during stance towards a net forward propulsion. We find that both a rotary force applied in the first half of stance and a radial force applied in the second half of stance, are capable of stabilizing running. Interestingly, these two forcing methods improve the motion stability in different ways. Rotary forcing first reduces then greatly increases the size of the stable parameter region when gradually increased. Radial forcing expands the stable parameter region, but only in a moderate way. Also, it is found that parameter region stabilized by rotary and radial forcing are largely complementary. Overall, rotary forcing can better stabilize running for both constant and event-based forcing functions that were attempted. This indicates that rotary forcing has an inherent capability of stabilizing running, even when minimal time-or-event-or-state feedback is present. Radial forcing, however, tends to be more energy efficient when compared to rotary forcing

Self-assembled morphologies of an amphiphilic Y-shaped weak polyelectrolyte in a thin film.

Science.gov (United States)

Mu, Dan; Li, Jian-Quan; Feng, Sheng-Yu

2017-11-29

Different from the self-assembly of neutral polymers, polyelectrolytes self-assemble into smaller aggregates with a more loosely assembled structure, which results from the repulsive forces acting between similar electrical compositions with the introduction of ions. The Y-shaped weak polyelectrolytes self-assemble into a core-shell type cylindrical structure with a hexagonal arrangement in a thin film, whose thickness is smaller than the gyration radius of the polymer chain. The corresponding formation mechanism consists of enrichment of the same components, adjustment of the shape of the aggregate, and the subsequent separation into individual aggregates. With the increase in the thickness of the thin film until it exceeds the gyration radius of the polymer chain, combined with the greater freedom of movement along the direction of thin film thickness, the self-assembled structure changes into a micellar structure. Under confinement, the repulsive force to the polymeric components is weakened by the repulsive forces among polyelectrolyte components with like charges, and this helps in generating aggregates with more uniform size and density distribution. In particular, when the repulsive force between the walls and the core forming components is greater than that between the walls and the shell forming components, such asymmetric confinement produces a crossed-cylindrical structure with nearly perpendicular arrangement of two cylinder arrays. Similarly, a novel three-crossed cylinder morphology is self-assembled upon removal of confinement.

TRIGA low enrichment fuel

International Nuclear Information System (INIS)

Gietzen, A.

1993-01-01

Sixty TRIGA reactors have been sold and the earliest of these are now passing twenty years of operation. All of these reactors use the uranium zirconium hydride fuel (UZrH) which provides certain unique advantages arising out of its large prompt negative temperature coefficient, very low fission product release, and high temperature capability. Eleven of these Sixty reactors are conversions from plate fuel to TRIGA fuel which were made as a result of these advantages. With only a few exceptions, TRIGA reactors have always used low-enriched uranium (LEU) fuel with an enrichment of 19.9%. The exceptions have either been converted from the standard low-enriched fuel to the 70% enriched FLIP fuel in order to achieve extended lifetime, or are higher powered reactors which were designed for long life using 93%-enriched uranium during the time when the use and export of highly enriched uranium (HEU) was not restricted. The advent of international policies focusing attention on nonproliferation and safeguards made the HEU fuels obsolete. General Atomic immediately undertook a development effort (nearly two years ago) in order to be in a position to comply with these policies for all future export sales and also to provide a low-enriched alternative to fully enriched plate-type fuels. This important work was subsequently partially supported by the U.S. Department of Energy. The laboratory and production tests have shown that higher uranium densities can be achieved to compensate for reducing the enrichment to 20%, and that the fuels maintain the characteristics of the very thoroughly proven standard TRIGA fuels. In May of 1978, General Atomic announced that these fuels were available for TRIGA reactors and for plate-type reactors with power levels up to 15 MW with General Atomic's standard commercial warranty

TRIGA low enrichment fuel

International Nuclear Information System (INIS)

Gietzen, A.

1993-01-01

Sixty TRIGA reactors have been sold and the earliest of these are now passing twenty years of operation. All of these reactors use the uranium-zirconium hydride fuel (UZrH) which provides certain unique advantages arising out of its large prompt negative temperature coefficient, very low fission product release, and high temperature capability. Eleven of these Sixty reactors are conversions from plate fuel to TRIGA fuel which were made as a result of these advantages. With only a few exceptions, TRIGA reactors have always used low-enriched-uranium (LEU) fuel with an enrichment of 19.9%. The exceptions have either been converted from the standard low-enriched fuel to the 70% enriched FLIP fuel in order to achieve extended lifetime, or are higher powered reactors which were designed for long life using 93%-enriched uranium during the time when the use and export of highly enriched uranium (HEU) was not restricted. The advent of international policies focusing attention on nonproliferation and safeguards made the HEU fuels obsolete. General Atomic immediately undertook a development effort (nearly two years ago) in order to be in a position to comply with these policies for all future export sales and also to provide a low-enriched alternative to fully enriched plate-type fuels. This important work was subsequently partially supported by the U.S. Department of Energy. The laboratory and production tests have shown that higher uranium densities can be achieved to compensate for reducing the enrichment to 20%, and that the fuels maintain the characteristics of the very thoroughly proven standard TRIGA fuels. In May of 1978, General Atomic announced that these fuels were available for TRIGA reactors and for plate-type reactors with power levels up to 15 MW with GA's standard commercial warranty

Fast radial basis functions for engineering applications

CERN Document Server

Biancolini, Marco Evangelos

2017-01-01

This book presents the first “How To” guide to the use of radial basis functions (RBF). It provides a clear vision of their potential, an overview of ready-for-use computational tools and precise guidelines to implement new engineering applications of RBF. Radial basis functions (RBF) are a mathematical tool mature enough for useful engineering applications. Their mathematical foundation is well established and the tool has proven to be effective in many fields, as the mathematical framework can be adapted in several ways. A candidate application can be faced considering the features of RBF:  multidimensional space (including 2D and 3D), numerous radial functions available, global and compact support, interpolation/regression. This great flexibility makes RBF attractive – and their great potential has only been partially discovered. This is because of the difficulty in taking a first step toward RBF as they are not commonly part of engineers’ cultural background, but also due to the numerical complex...

On improved confinement in mirror plasmas by a radial electric field

Science.gov (United States)

Ågren, O.; Moiseenko, V. E.

2017-11-01

A weak radial electric field can suppress radial excursions of a guiding center from its mean magnetic surface. The physical origin of this effect is the smearing action by a poloidal E × B rotation, which tend to cancel out the inward and outward radial drifts. A use of this phenomenon may provide larger margins for magnetic field shaping with radial confinement of particles maintained in the collision free idealization. Mirror fields, stabilized by a quadrupolar field component, are of particular interest for their MHD stability and the possibility to control the quasi neutral radial electric field by biased potential plates outside the confinement region. Flux surface footprints on the end tank wall have to be traced to avoid short-circuiting between biased plates. Assuming a robust biasing procedure, moderate voltage demands for the biased plates seems adequate to cure even the radial excursions of Yushmanov ions which could be locally trapped near the mirrors. Analytical expressions are obtained for a magnetic quadrupolar mirror configuration which possesses minimal radial magnetic drifts in the central confinement region. By adding a weak controlled radial quasi-neutral electric field, the majority of gyro centers are predicted to be forced to move even closer to their respective mean magnetic surface. The gyro center radial coordinate is in such a case an accurate approximation for a constant of motion. By using this constant of motion, the analysis is in a Vlasov description extended to finite β. A correspondence between that Vlasov system and a fluid description with a scalar pressure and an electric potential is verified. The minimum B criterion is considered and implications for flute mode stability in the considered magnetic field is analyzed. By carrying out a long-thin expansion to a higher order, the validity of the calculations are extended to shorter and more compact device designs.

Spent nuclear fuel assembly storage vessel

International Nuclear Information System (INIS)

Yagishita, Takuya

1998-01-01

The vessel of the present invention promotes an effect of removing after heat of spent nuclear fuel assemblies so as not to give force to the storage vessel caused by expansion of heat removing partitioning plates. Namely, the vessel of the present invention comprises a cylinder body having closed upper and lower portions and a plurality of heat removing partitioning cylinders disposed each at a predetermined interval in the circumferential direction of the above-mentioned cylinder body. The heat removing partitioning cylinders comprises (1) first heat removing partitioning plates extended in the radial direction of the cylinder body and opposed at a predetermined gap in the circumferential direction of the cylinder body, and having the base ends on the side of the inner wall of the cylinder body being secured to the inner wall of the cylinder body and (2) a second heat removing plate for connecting the top ends of both opposed heat removing partitioning plates on the central side of the cylinder body with each other. Spent nuclear fuel assemblies are contained in a plurality of closed spaces surrounded by the first heat removing partitioning plates and the second heat removing partitioning plate. With such constitution, since after heat is partially transferred from the heat removing partitioning plates to the cylindrical body directly by heat conduction, the heat removing effect can be promoted compared with the prior art. (I.S.)

Turkey's regulatory plans for high enriched to low enriched conversion of TR-2 reactor core

International Nuclear Information System (INIS)

Guelol Oezdere, Oya

2003-01-01

Turkey is a developing country and has three nuclear facilities two of which are research reactors and one pilot fuel production plant. One of the two research reactors is TR-2 which is located in Cekmece site in Istanbul. TR-2 Reactor's core is composed of both high enriched and low enriched fuel and from high enriched to low enriched core conversion project will take place in year 2005. This paper presents the plans for drafting regulations on the safety analysis report updates for high enriched to low enriched core conversion of TR-2 reactor, the present regulatory structure of Turkey and licensing activities of nuclear facilities. (author)

Radial Transport and Meridional Circulation in Accretion Disks

Energy Technology Data Exchange (ETDEWEB)

Philippov, Alexander A. [Department of Astrophysical Sciences, Princeton University, Ivy Lane, Princeton, NJ 08540 (United States); Rafikov, Roman R., E-mail: sashaph@princeton.edu [Institute for Advanced Study, Einstein Drive, Princeton, NJ 08540 (United States)

2017-03-10

Radial transport of particles, elements and fluid driven by internal stresses in three-dimensional (3D) astrophysical accretion disks is an important phenomenon, potentially relevant for the outward dust transport in protoplanetary disks, origin of the refractory particles in comets, isotopic equilibration in the Earth–Moon system, etc. To gain better insight into these processes, we explore the dependence of meridional circulation in 3D disks with shear viscosity on their thermal stratification, and demonstrate a strong effect of the latter on the radial flow. Previous locally isothermal studies have normally found a pattern of the radial outflow near the midplane, switching to inflow higher up. Here we show, both analytically and numerically, that a flow that is inward at all altitudes is possible in disks with entropy and temperature steeply increasing with height. Such thermodynamic conditions may be typical in the optically thin, viscously heated accretion disks. Disks in which these conditions do not hold should feature radial outflow near the midplane, as long as their internal stress is provided by the shear viscosity. Our results can also be used for designing hydrodynamical disk simulations with a prescribed pattern of the meridional circulation.

Detonation in supersonic radial outflow

KAUST Repository

Kasimov, Aslan R.; Korneev, Svyatoslav

2014-01-01

We report on the structure and dynamics of gaseous detonation stabilized in a supersonic flow emanating radially from a central source. The steady-state solutions are computed and their range of existence is investigated. Two-dimensional simulations

High enrichment to low enrichment core's conversion. Technical securities

International Nuclear Information System (INIS)

Abbate, P.; Madariaga, M.R.

1990-01-01

This work presents the fulfillment of the technical securities subscribed by INVAP S.E. for the conversion of a high enriched uranium core. The reactor (of 5 thermal Mw), built in the 50's and 60's, is of the 'swimming pool' type, with light water and fuel elements of the curve plates MTR type, enriched at 93.15 %. These are neutronic and thermohydraulic securities. (Author) [es

Beta activity of enriched uranium

International Nuclear Information System (INIS)

Nambiar, P.P.V.J.; Ramachandran, V.

1975-01-01

Use of enriched uranium as reactor fuel necessitates its handling in various forms. For purposes of planning and organising radiation protection measures in enriched uranium handling facilities, it is necessary to have a basic knowledge of the radiation status of enriched uranium systems. The theoretical variations in beta activity and energy with U 235 enrichment are presented. Depletion is considered separately. Beta activity build up is also studied for two specific enrichments, in respect of which experimental values for specific alpha activity are available. (author)

THE OCCURRENCE OF THE RADIAL CLUB HAND IN CHILDREN WITH DIFFERENT SYNDROMES

Directory of Open Access Journals (Sweden)

Sergey Ivanovich Golyana

2013-03-01

Full Text Available Radial club hand is a developmental anomaly of the upper extremity, being characterized as a longitudinal underdevelopment of a forearm and a hand on the radial surface, consisting in a hypo-/ aplazy radial bone and the thumb of various degree of expressiveness. Characteristic symptoms of this developmental anomaly are: shortening and bow-shaped curvature of a forearm, palmar and radial deviation of a hand, underdevelopment of the thumb from its proximal departments and structures, anomaly of development of three-phalanx fingers of a hand (is more often than the 2-4th, violation of a cosmetic condition and functionality of the affected segment. From 2000 for 2012 in FSI SRICO n.a. H.Turner examination and treatment of 23 children with various syndromes at which the radial club hand was revealed are conducted. The main syndromes at which it is revealed radial club hand - Holt-Orama syndrome, TAR- syndrome and VACTERL syndrome. Tactics and techniques of surgical treatment of a radial club hand it various syndromes most often don’t differ from treatment of other types of a radial club hand though demand an individual approach depending on severity and a type of deformation of the upper extremity.

Enriching step-based product information models to support product life-cycle activities

Science.gov (United States)

Sarigecili, Mehmet Ilteris

The representation and management of product information in its life-cycle requires standardized data exchange protocols. Standard for Exchange of Product Model Data (STEP) is such a standard that has been used widely by the industries. Even though STEP-based product models are well defined and syntactically correct, populating product data according to these models is not easy because they are too big and disorganized. Data exchange specifications (DEXs) and templates provide re-organized information models required in data exchange of specific activities for various businesses. DEXs show us it would be possible to organize STEP-based product models in order to support different engineering activities at various stages of product life-cycle. In this study, STEP-based models are enriched and organized to support two engineering activities: materials information declaration and tolerance analysis. Due to new environmental regulations, the substance and materials information in products have to be screened closely by manufacturing industries. This requires a fast, unambiguous and complete product information exchange between the members of a supply chain. Tolerance analysis activity, on the other hand, is used to verify the functional requirements of an assembly considering the worst case (i.e., maximum and minimum) conditions for the part/assembly dimensions. Another issue with STEP-based product models is that the semantics of product data are represented implicitly. Hence, it is difficult to interpret the semantics of data for different product life-cycle phases for various application domains. OntoSTEP, developed at NIST, provides semantically enriched product models in OWL. In this thesis, we would like to present how to interpret the GD & T specifications in STEP for tolerance analysis by utilizing OntoSTEP.

The effect of radial migration on galactic disks

International Nuclear Information System (INIS)

Vera-Ciro, Carlos; D'Onghia, Elena; Navarro, Julio; Abadi, Mario

2014-01-01

We study the radial migration of stars driven by recurring multi-arm spiral features in an exponential disk embedded in a dark matter halo. The spiral perturbations redistribute angular momentum within the disk and lead to substantial radial displacements of individual stars, in a manner that largely preserves the circularity of their orbits and that results, after 5 Gyr (∼40 full rotations at the disk scale length), in little radial heating and no appreciable changes to the vertical or radial structure of the disk. Our results clarify a number of issues related to the spatial distribution and kinematics of migrators. In particular, we find that migrators are a heavily biased subset of stars with preferentially low vertical velocity dispersions. This 'provenance bias' for migrators is not surprising in hindsight, for stars with small vertical excursions spend more time near the disk plane, and thus respond more readily to non-axisymmetric perturbations. We also find that the vertical velocity dispersion of outward migrators always decreases, whereas the opposite holds for inward migrators. To first order, newly arrived migrators simply replace stars that have migrated off to other radii, thus inheriting the vertical bias of the latter. Extreme migrators might therefore be recognized, if present, by the unexpectedly small amplitude of their vertical excursions. Our results show that migration, understood as changes in angular momentum that preserve circularity, can strongly affect the thin disk, but cast doubts on models that envision the Galactic thick disk as a relic of radial migration.

Rayleigh-Taylor instability of cylindrical jets with radial motion

International Nuclear Information System (INIS)

Chen, X.M.; Schrock, V.E.; Peterson, P.F.

1997-01-01

Rayleigh-Taylor instability of an interface between fluids with different densities subjected to acceleration normal to itself has interested researchers for almost a century. The classic analyses of a flat interface by Rayleigh and Taylor have shown that this type of instability depends on the direction of acceleration and the density differences of the two fluids. Plesset later analyzed the stability of a spherically symmetric flows (and a spherical interface) and concluded that the instability also depends on the velocity of the interface as well as the direction and magnitude of radial acceleration. The instability induced by radial motion in cylindrical systems seems to have been neglected by previous researchers. This paper analyzes the Rayleigh-Taylor type of instability for a cylindrical surface with radial motions. The results of the analysis show that, like the spherical case, the radial velocity also plays an important role. As an application, the example of a liquid jet surface in an Inertial Confinement Fusion (ICF) reactor design is analyzed. (orig.)

A Novel Integrated Structure with a Radial Displacement Sensor and a Permanent Magnet Biased Radial Magnetic Bearing

Directory of Open Access Journals (Sweden)

Jinji Sun

2014-01-01

Full Text Available In this paper, a novel integrated structure is proposed in order to reduce the axial length of the high speed of a magnetically suspended motor (HSMSM to ensure the maximum speed, which combines radial displacement sensor probes and the permanent magnet biased radial magnetic bearing in HSMSM. The sensor probes are integrated in the magnetic bearing, and the sensor preamplifiers are placed in the control system of the HSMSM, separate from the sensor probes. The proposed integrated structure can save space in HSMSMs, improve the working frequency, reduce the influence of temperature on the sensor circuit, and improve the stability of HSMSMs.

Radial-piston pump for drive of test machines

Science.gov (United States)

Nizhegorodov, A. I.; Gavrilin, A. N.; Moyzes, B. B.; Cherkasov, A. I.; Zharkevich, O. M.; Zhetessova, G. S.; Savelyeva, N. A.

2018-01-01

The article reviews the development of radial-piston pump with phase control and alternating-flow mode for seismic-testing platforms and other test machines. The prospects for use of the developed device are proved. It is noted that the method of frequency modulation with the detection of the natural frequencies is easily realized by using the radial-piston pump. The prospects of further research are given proof.

The substitution technique applied for determining the parameters of a reactor with void

International Nuclear Information System (INIS)

Zdravkovic, Z.; Ivkovic, M.; Sotic, O.

1965-12-01

The material buckling of D 2 O-enriched uranium (2% U 235 ) system with voided fuel assembly has been determined by means of the substitution technique. The experiments were carried out in the zero power reactor RB for various lattice pitches, namely: 11.3 cm, 14 cm, and 16 cm. As a reference core the D 2 O-enriched uranium (2%U 235 ) ) system without voids has been used. The results have been analysed by the modified one-group perturbation theory using three regions in a substituted reactor (test, intermediate and reference). The perturbations in axial and radial directions due to voided fuel assembly, are also considered. In order to determine the change between diffusion coefficients in various regions (δD), additional experiments with single test fuel element were made. The lattice pitches of reference and test regions were the same. According to the algorithms given in appendices, the programmes for the ZUSE Z-23 digital computer were made. These programmes include evaluations of the change between diffusion coefficients (δD), and the axial buckling of the test lattice (author)

SpicyNodes Radial Map Engine

Science.gov (United States)

Douma, M.; Ligierko, G.; Angelov, I.

2008-10-01

The need for information has increased exponentially over the past decades. The current systems for constructing, exploring, classifying, organizing, and searching information face the growing challenge of enabling their users to operate efficiently and intuitively in knowledge-heavy environments. This paper presents SpicyNodes, an advanced user interface for difficult interaction contexts. It is based on an underlying structure known as a radial map, which allows users to manipulate and interact in a natural manner with entities called nodes. This technology overcomes certain limitations of existing solutions and solves the problem of browsing complex sets of linked information. SpicyNodes is also an organic system that projects users into a living space, stimulating exploratory behavior and fostering creative thought. Our interactive radial layout is used for educational purposes and has the potential for numerous other applications.

Isotope enrichment

International Nuclear Information System (INIS)

Garbuny, M.

1979-01-01

The invention discloses a method for deriving, from a starting material including an element having a plurality of isotopes, derived material enriched in one isotope of the element. The starting material is deposited on a substrate at less than a critical submonatomic surface density, typically less than 10 16 atoms per square centimeter. The deposit is then selectively irradiated by a laser (maser or electronic oscillator) beam with monochromatic coherent radiation resonant with the one isotope causing the material including the one istope to escape from the substrate. The escaping enriched material is then collected. Where the element has two isotopes, one of which is to be collected, the deposit may be irradiated with radiation resonant with the other isotope and the residual material enriched in the one isotope may be evaporated from the substrate and collected

Modelling and analysis of radial thermal stresses and temperature ...

African Journals Online (AJOL)

A theoretical investigation has been undertaken to study operating temperatures, heat fluxes and radial thermal stresses in the valves of a modern diesel engine with and without air-cavity. Temperatures, heat fluxes and radial thermal stresses were measured theoretically for both cases under all four thermal loading ...

Laser and gas centrifuge enrichment

Energy Technology Data Exchange (ETDEWEB)

Heinonen, Olli [Senior Fellow, Belfer Center for Science and International Affairs, Harvard Kennedy School, Cambridge, Massachusetts (United States)

2014-05-09

Principles of uranium isotope enrichment using various laser and gas centrifuge techniques are briefly discussed. Examples on production of high enriched uranium are given. Concerns regarding the possibility of using low end technologies to produce weapons grade uranium are explained. Based on current assessments commercial enrichment services are able to cover the global needs of enriched uranium in the foreseeable future.

A novel integrated 4-DOF radial hybrid magnetic bearing for MSCMG

Energy Technology Data Exchange (ETDEWEB)

Jinji, Sun; Ziyan, Ju [School of Instrumentation Science & Opto-electronics Engineering, Beijing University of Aeronautics and Astronautics, Science and Technology on Inertial Laboratory, Beijing 100191 (China); Weitao, Han, E-mail: hanweitaotao@163.com [CRRC Qingdao Sifang CO., LTD, Qingdao 266111 (China); Gang, Liu [School of Instrumentation Science & Opto-electronics Engineering, Beijing University of Aeronautics and Astronautics, Science and Technology on Inertial Laboratory, Beijing 100191 (China)

2017-01-01

This paper proposes a novel integrated radial hybrid magnetic bearing (RHMB) for application with the small-sized magnetically suspended control moment gyroscope (MSCMG), which can control four degrees of freedom (4-DOFs), including two radial translational DOFs and two radial tilting DOFs, and provide the axial passive resilience. The configuration and working principle of the RHMB are introduced. Mathematical models of radial force, axial resilience and moment are established by using equivalent magnetic circuit method (EMCM), from which the radial force–radial displacement, radial force–current relationships are derived, as well as axial resilience–axial displacement, moment–tilting angle and moment–current. Finite element method (FEM) is also applied to analyze the performance and characteristics of the RHMB. The analysis results are in good agreement with that calculated by the EMCM, which is helpful in designing, optimizing and controlling the RHMB. The comparisons between the performances of the integrated 4-DOF RHMB and the traditional 4-DOF RHMB are made. The contrast results indicate that the proposed integrated 4-DOF RHMB possesses better performance compared to the traditional structure, such as copper loss, current stiffness, and tilting current stiffness. - Highlights: • An integrated 4-DOF RHMB is proposed for the small-sized MSCMG. • The 4-DOF RHMB has good linear force–displacement and force–current characteristics. • The RHMB has good linear moment–current and the moment–tilting angle characteristic.

Enrichment technology. Dependable vendor of gas centrifuges; Enrichment Technology. Zuverlaessiger Lieferant von Gaszentrifugen

Energy Technology Data Exchange (ETDEWEB)

Anon.

2011-10-15

Enrichment Technology is an innovative, high-tech company that develops, manufactures and installs gas centrifuges for enriching uranium. In addition, Enrichment Technology designs enrichment plants that use gas centrifuge technology. This technology offers the most efficient and cost-effective method for enriching uranium yet: high-performance, safe technology that dominates the market with a global share of 45 percent. A determining factor in Enrichment Technology's success is its mission: supplying its customers with safe, reliable technology. Production of the centrifuges requires versatile know-how and collaboration between different departments as well as interdisciplinary teams at the various sites. More than 2000 operators at 8 sites in 5 countries contribute their individual knowledge and personal skills in order to produce this exceptional technology. The head office is in Beaconsfield near London and the operational headquarters are in Almelo in the Netherlands. There are other sites in Germany (Juelich und Gronau), Great Britain (Capenhurst) as well as project sites in the USA and France. Capenhurst is where experienced engineers design new enrichment plants and organise their construction. Centrifuge components are manufactured in Almelo and Juelich, while the pipework needed to connect up the centrifuges is produced at the site in Gronau. In Juelich, highly qualified scientists in interdisciplinary teams are continuously researching ways of improving the current centrifuges. Communication between specialists in the fields of chemistry, physics and engineering forms the basis for the company's success and the key to extending this leading position in the global enrichment market. (orig.)

Burnup verification measurements on spent fuel assemblies at Arkansas Nuclear One

International Nuclear Information System (INIS)

Ewing, R.I.

1995-01-01

Burnup verification measurements have been performed using the Fork system at Arkansas Nuclear One, Units 1 and 2, operated by Energy Operations, Inc. Passive neutron and gamma-ray measurements on individual spent fuel assemblies were correlated with the reactor records for burnup, cooling time, and initial enrichment. The correlation generates an internal calibration for the system in the form of a power law determined by a least squares fit to the neutron data. The values of the exponent in the power laws were 3.83 and 4.35 for Units 1 and 2, respectively. The average deviation of the reactor burnup records from the calibration determined from the measurements is a measure of the random error in the burnup records. The observed average deviations were 2.7% and 3.5% for assemblies at Units 1 and 2, respectively, indicating a high degree of consistency in the reactor records. Two non-standard assemblies containing neutron sources were studied at Unit 2. No anomalous measurements were observed among the standard assemblies at either Unit. The effectiveness of the Fork system for verification of reactor records is due to the sensitivity of the neutron yield to burnup, the self-calibration generated by a series of measurements, the redundancy provided by three independent detection systems, and the operational simplicity and flexibility of the design

Predicting fissile content of spent nuclear fuel assemblies with the Passive Neutron Albedo Reactivity technique and Monte Carlo code emulation

International Nuclear Information System (INIS)

Conlin, Jeremy Lloyd; Tobin, Stephen J.

2011-01-01

There is a great need in the safeguards community to be able to nondestructively quantify the mass of plutonium of a spent nuclear fuel assembly. As part of the Next Generation of Safeguards Initiative, we are investigating several techniques, or detector systems, which, when integrated, will be capable of quantifying the plutonium mass of a spent fuel assembly without dismantling the assembly. This paper reports on the simulation of one of these techniques, the Passive Neutron Albedo Reactivity with Fission Chambers (PNAR-FC) system. The response of this system over a wide range of spent fuel assemblies with different burnup, initial enrichment, and cooling time characteristics is shown. A Monte Carlo method of using these modeled results to estimate the fissile content of a spent fuel assembly has been developed. A few numerical simulations of using this method are shown. Finally, additional developments still needed and being worked on are discussed. (author)

MRI of radial displacement of the meniscus in the knee

International Nuclear Information System (INIS)

Chen Jian; Lv Houshan; Lao Shan; Guan Zhenpeng; Hong Nan; Liang Hao

2006-01-01

Objective: To describe the phenomenon of radial displacement of the meniscus of the knees in the study population with MR imaging, and to establish MRI diagnostic criteria for radial displacement of the meniscus and displacement index. Methods: MR signs of radial displacement of the meniscus were evaluated retrospectively in 398 patients with knee symptoms who were examined with non- weight bearing MR images from Jan. 2000 to Feb. 2004. The patients younger than 18 years old, with joint effusion or serious arthropathy were excluded and 312 patients were eligible to be enrolled in this study. The criterion for radial displacement of the meniscus was defined as the location of the edge of meniscal body beyond the femoral and tibial outer border line. A displacement index, defined as the ratio of meniscal overhang to meniscal width, was used to quantify meniscal displacement. Results: The prevalence of radial displacement of the meniscus was 16.7% (52/312) and 13.9% (21/151) in right knee and 19.3% (31/161 )in left knee, respectively. There was no significant difference between left and right knee (χ 2 =1.60, P>0.05) and the ratio between medial and lateral meniscus was 7.8:1. The average displacement index was 0.54±0.24. The displacement indices were significant higher in older group (F=3.63, P<0.05). The incidence and indices of radial displacement of the meniscus for patients under or above 50 year older were 12.0%(17/142), 0.46±0.22 and 20.6% (35/170), 0.64±0.20, respectively. Difference was highly significant (t=0.84, P<0.01). Conclusion: It was concluded that radial displacement of the meniscus in knees was not a rare finding with MR imaging in patients with knee symptoms. The incidence increased in older age group. Further investigations were recommended to understand the etiology and clinical significance of the phenomenon of radial displacement of the meniscus. (authors)

76 FR 11523 - Atomic Safety and Licensing Board; AREVA Enrichment Services, LLC (Eagle Rock Enrichment Facility...

Science.gov (United States)

2011-03-02

... and Licensing Board; AREVA Enrichment Services, LLC (Eagle Rock Enrichment Facility); Notice of... Governmental Entities Regarding Environmental Portion of Enrichment Facility Licensing Proceeding February 24.... White. In this 10 CFR part 70 proceeding regarding the request of applicant AREVA Enrichment Services...

76 FR 34103 - In the Matter of Areva Enrichment Services, LLC (Eagle Rock Enrichment Facility); Notice of...

Science.gov (United States)

2011-06-10

.... 10-899-02-ML-BD01] In the Matter of Areva Enrichment Services, LLC (Eagle Rock Enrichment Facility...'' portion of this proceeding regarding the December 2008 application by AREVA Enrichment Services, LLC (AES... gas centrifuge uranium enrichment facility--denoted as the Eagle Rock Enrichment Facility (EREF)--in...

Vortex Whistle in Radial Intake

National Research Council Canada - National Science Library

Tse, Man-Chun

2004-01-01

In a radial-to-axial intake with inlet guide vanes (IGV) at the entry, a strong flow circulation Gamma can be generated from the tangential flow components created by the IGVs when their setting exceed about halfclosing (approx. 45 deg...

Radial force measurement of endovascular stents: Influence of stent design and diameter.

Science.gov (United States)

Matsumoto, Takuya; Matsubara, Yutaka; Aoyagi, Yukihiko; Matsuda, Daisuke; Okadome, Jun; Morisaki, Koichi; Inoue, Kentarou; Tanaka, Shinichi; Ohkusa, Tomoko; Maehara, Yoshihiko

2016-04-01

Angioplasty and endovascular stent placement is used in case to rescue the coverage of main branches to supply blood to brain from aortic arch in thoracic endovascular aortic repair. This study assessed mechanical properties, especially differences in radial force, of different endovascular and thoracic stents. We analyzed the radial force of three stent models (Epic, E-Luminexx and SMART) stents using radial force-tester method in single or overlapping conditions. We also analyzed radial force in three thoracic stents using Mylar film testing method: conformable Gore-TAG, Relay, and Valiant Thoracic Stent Graft. Overlapping SMART stents had greater radial force than overlapping Epic or Luminexx stents (P stents was greater than that of all three endovascular stents (P stents, site of deployment, and layer characteristics. In clinical settings, an understanding of the mechanical characteristics, including radial force, is important in choosing a stent for each patient. © The Author(s) 2015.

PROLONGED RADIAL ARTERY SPASM IN THE CATHETERIZATION LABORATORY - RELIEF BY PHARMACOLOGICAL INTERVENTION

Directory of Open Access Journals (Sweden)

Krishna Kumar

2010-11-01

Full Text Available Radial spasm is often very prolonged and painful to the patient. Here, we describe a novel way to deal with the same. The total spasm lasted over 4 hours. A 3.4 6 JR catheter was introduced via the femoral route and papav arine one ampoule was injected directly into the right subclavian artery. After about 10 min we were able to pull out the radial catheter. Radial angiography is a simple procedure with reportedly less complications 1,2. How ever ,it has one major complication radial spasm. We describe here a patient with radial spasm that persisted for more than 2 hours and how we dealt with it.

Rotary and radial forcing effects on center-of-mass locomotion dynamics

International Nuclear Information System (INIS)

Shen, Z H; Larson, P L; Seipel, J E

2014-01-01

Rotary and radial forcing are two common actuation methods for legged robots. However, these two orthogonal methods of center-of-mass (CoM) forcing have not been compared as potentially alternative strategies of actuation. In this paper, we compare the CoM stability and energetics of running with rotary and radial actuation through the simulation of two models: the rotary-forced spring-loaded inverted pendulum (rotary-forced-SLIP), and the radially-forced-SLIP. We model both radial and rotary actuation in the simplest way, applying them as a constant force during the stance portion of the gait. A simple application of constant rotary forcing throughout stance is capable of producing fully-asymptotically stable motion; however, a similarly constant application of radial forcing throughout the stance is not capable of producing stable solutions. We then allow both the applied rotary and radial forcing functions to turn on or off based on the occurrence of the mid-stance event, which breaks the symmetry of actuation during stance towards a net forward propulsion. We find that both a rotary force applied in the first half of stance and a radial force applied in the second half of stance, are capable of stabilizing running. Interestingly, these two forcing methods improve the motion stability in different ways. Rotary forcing first reduces then greatly increases the size of the stable parameter region when gradually increased. Radial forcing expands the stable parameter region, but only in a moderate way. Also, it is found that parameter region stabilized by rotary and radial forcing are largely complementary. Overall, rotary forcing can better stabilize running for both constant and event-based forcing functions that were attempted. This indicates that rotary forcing has an inherent capability of stabilizing running, even when minimal time-or-event-or-state feedback is present. Radial forcing, however, tends to be more energy efficient when compared to rotary forcing

Enrichment plants. A survey of major new uranium enriching projects

International Nuclear Information System (INIS)

Kovan, D.

1976-01-01

The work enrichment situation is reported. The development of enrichment in the U.S. and in Europe is outlined. A brief description is given of the technology of separation by diffusion and by centrifugation and the advantages and disadvantages of the two processes are compared. Finally the supply and demand situation is briefly considered. (U.K.)

The First Experience of Triple Nerve Transfer in Proximal Radial Nerve Palsy.

Science.gov (United States)

Emamhadi, Mohammadreza; Andalib, Sasan

2018-01-01

Injury to distal portion of posterior cord of brachial plexus leads to palsy of radial and axillary nerves. Symptoms are usually motor deficits of the deltoid muscle; triceps brachii muscle; and extensor muscles of the wrist, thumb, and fingers. Tendon transfers, nerve grafts, and nerve transfers are options for surgical treatment of proximal radial nerve palsy to restore some motor functions. Tendon transfer is painful, requires a long immobilization, and decreases donor muscle strength; nevertheless, nerve transfer produces promising outcomes. We present a patient with proximal radial nerve palsy following a blunt injury undergoing triple nerve transfer. The patient was involved in a motorcycle accident with complete palsy of the radial and axillary nerves. After 6 months, on admission, he showed spontaneous recovery of axillary nerve palsy, but radial nerve palsy remained. We performed triple nerve transfer, fascicle of ulnar nerve to long head of the triceps branch of radial nerve, flexor digitorum superficialis branch of median nerve to extensor carpi radialis brevis branch of radial nerve, and flexor carpi radialis branch of median nerve to posterior interosseous nerve, for restoration of elbow, wrist, and finger extensions, respectively. Our experience confirmed functional elbow, wrist, and finger extensions in the patient. Triple nerve transfer restores functions of the upper limb in patients with debilitating radial nerve palsy after blunt injuries. Copyright © 2017 Elsevier Inc. All rights reserved.

Field test and evaluation of the IAEA coincidence collar for the measurement of unirradiated BWR fuel assemblies

International Nuclear Information System (INIS)

Menlove, H.O.; Keddar, A.

1982-12-01

The neutron coincidence counter has been field tested and evaluated for the measurement of boiling-water-reactor (BWR) fuel assemblies at the ASEA-ATOM Fuel Fabrication Facility. The system measures the 235 U content per unit length of full fuel assemblies using neutron interrogation and coincidence counting. The 238 U content is measured in the passive mode without the AmLi neutron interrogatioin source. The field tests included both standard production movable fuel rods to investigate enrichment and absorber variations. Results gave a response standard deviation of 0.9% for the active case and 2.1% for the passive case in 1000-s measurement times. 10 figures, 2 tables