Maximum Thermal Flux Research Articles

Effect of Weave and Weft Type on Mechanical and Comfort Properties of Hemp-Linen Fabrics.

In this study the influence of fabric weave (plain, twill, and panama) and weft type (flax and hemp yarns) on selected mechanical and comfort properties of six fabrics was analyzed. The results showed that tear and abrasion properties were most affected by the weave. The tensile properties of the linen fabrics were not significantly different when weft hemp yarns were used instead of flax. Fabrics with the same weave seemed to be equally resilient to abrasion regardless of the type of weft. By contrast, the hemp weft yarns favorized the physical and comfort properties of the investigated fabrics. For the same weave, the hemp-linen fabrics were slightly lighter and exhibited lower bulk density, significantly larger air permeability, and improved moisture management properties. Although the results of maximum thermal flux (Qmax) suggested a cooler sensation of the linen fabrics with panama and twill, the hemp-linen fabric with a plain weave seemed to be the optimal choice when a cool sensation was desired. Higher thermal conductivity values also suggested slightly better heat transfer properties of the hemp-linen fabrics, and these were significantly influenced by the weave type. The results clearly indicated the advantages of using hemp for improving physical and specific comfort properties of linen fabrics.

Multi-scale microstructure manipulation of an additively manufactured CoCrNi medium entropy alloy for superior mechanical properties and tunable mechanical anisotropy

Laser beam powder bed fusion (PBF-LB) additive manufacturing (AM) technology has become a versatile tool for producing new microstructures in metal components, offering novel mechanical properties for different applications. In this work, enhanced ductility (∼55% elongation) and tunable mechanical anisotropy (ratio of ductility along vertical to horizontal orientation from ∼0.2 to ∼1) were achieved for a CoCrNi medium entropy alloy (MEA) by multi-scale synergistic microstructure manipulation (i.e., melt pool boundary, grain morphology and crystallographic texture) through adjusting key PBF-LB processing parameters (e.g., laser power and scan speed). By increasing the volumetric energy density (VED) from 68.3 to 144 J/mm3, the melt pool size enlarges, and the crystallographic texture transitions from <100>//BD to <110>//BD due to the maximum thermal flux direction changing for different melt pool dimensions, which affects the proportion of grains that have different growth directions. Moreover, excellent mechanical properties of 890 MPa ultimate tensile strength and ∼55% elongation to failure can be achieved for loading perpendicular to the build direction with only a ∼15% reduction in properties for loading along the build direction. The superior combination of mechanical properties is achieved by processing parameter-controlled strengthening of melt pool boundary interfaces, heterogeneous deformation induced strengthening through bimodal grain structures, and favorable grain orientation for dislocation slip and twinning formation, which was significantly more activated in the <110>//BD texture than in the <100>//BD. This study offers new insights into achieving multi-scale synergistic microstructure manipulation via PBF-LB for desired strength, ductility, and mechanical anisotropy in a CoCrNi MEA.

Assessment of comfort property of jute-cotton blend denim trousers after repetitive laundering

In this analysis jute cotton blend denims performance at enzyme wash to 35 times respective laundering, in friction and roughness properties where 5 times laundered sample showed smoother surface in both inner and outer surface. In bending properties, 5 times laundered sample was found second lowest Bending Average Rigidity (BAR) and Bending Work (BW). In compression, highest Compression Average Rigidity (CAR), Recovery Average Rigidity(RAR) and less energy requirement Compression work (CW) to compress in both inner and outer surface was found in 5 times laundered sample. In Sensorial Comfort Aspect (SCA), 5 times laundered sample was found smoother surface property, quite higher bending property and higher compression average rigidity. Therefore, 5 times laundered sample was good sensorial comfort among all the samples. In Maximum Thermal Flux (MTF), 20 times laundered sample gives most warm feeling at inner surface because it provides minimum heat flux which ensure warmness in skin-garment micro environment. For thermo physiological comfort aspect, only 20 times repetitive laundered sample found maximum thermo physiological comfort among all samples. Thus, Jute-Cotton blend denims performance at enzyme wash to 35 times repetitive laundering up to 20 times repetitive laundering sample reported good results and the protruding fibre of jute was visible excessively on the surface.

Improvement of a design of the molten salt fast reactor

This study provides an improved design of molten salt fast reactors to reduce the maximum magnitude velocity in the core. The results indicated that the improved design reduces the magnitude velocity with acceptable hot spot temperature where the maximum magnitude velocity is nearly equal to 4.59 m/s and the temperature of the hot spot is greater than the outlet temperature by 9°. The performed analyses are used to adjust the amount of fissile material started for keff = 1. Neutron flux and delayed Neutron distribution study where the position of the maximum Neutron flux is in the center and the position of the maximum thermal flux is in a fertile blanket. Simulations of unprotected loss of heat sink, unprotected loss of flow, unprotected transient overpower, and Unprotected over-cooling are performed and discussed with new geometry. Due to the negative reactive feedback, it is clear that MSFR has a high safety potential.

Multiobjective Core Reloading Pattern Optimization of PARR-1 Using Modified Genetic Algorithm Coupled with Monte Carlo Methods

In order to maximize both the life cycle and efficiency of a reactor core, it is essential to find the optimum loading pattern. In the case of research reactors, a loading pattern can also be optimized for flux at an irradiation site. Therefore, the development of a general-use methodology for core loading optimization would be very valuable. In this paper, general-use multiobjective core reloading pattern optimization is performed using modified genetic algorithms (MGA). The developed strategy can be applied for the constrained optimization of research and power reactor cores. For an optimal reactor core reloading design strategy, an intelligent technique GA is coupled with the Monte Carlo (MC) code SuperMC developed by the FDS team in China for nuclear reactor physics calculations. An optimal loading pattern can be depicted as a configuration that has the maximum keff and maximum thermal fluxes in the core of the given fuel inventory keeping in view the safety constraints such as limitation on power peaking factor. The optimized loading patterns for Pakistan Research Reactor-1 (PARR-1) have been recommended using the implemented strategy by considering the constraint optimization, i.e., to maximize the keff or maximum thermal neutron flux while maintaining low power peaking factor. It has been observed that the developed intelligent strategy performs these tasks with a reasonable computational cost.

Conceptual Design of a 10 MW Multipurpose Research Reactor Using VVR-KN Fuel

The paper presents a conceptual design of a 10 MW multipurpose nuclear research reactor (MPRR) loaded with the low-enriched uranium (LEU) VVR-KN fuel type. Neutronics and burnup calculations have been performed using the REBUS-MCNP6 linkage system code and the ENDF/B-VII.0 data library. The core consists of 36 fuel assemblies: 27 standard fuel assemblies and 9 control fuel assemblies with the uranium density of 2.8 gU/cm3 and the 235U enrichment of 19.75 wt.%. The cycle length of the core is 86 effective full-power days with the excess reactivity of 9600 and 1039 pcm at the beginning of cycle and the end of cycle, respectively. The highest power rate and the highest discharged burnup of fuel assembly are 393.49 kW and 56.74% loss of 235U, respectively. Thermal hydraulics analysis has also been conducted using the PLTEMP4.2 code for evaluating the safety parameters at a steady state of the hottest channel. The maximum temperatures of coolant and fuel cladding are 66.0°C and 83.0°C, respectively. This value is lower than the design limit of 98°C for cladding temperature. Thermal fluxes at the vertical irradiation channels and the horizontal beam ports have been evaluated. The maximum thermal fluxes of 2.5 × 1014 and 8.9 ×1013 n·cm−2·s−1 are found at the neutron trap and the beryllium reflector, respectively.

Design of infrared radiation diagnostic calorimeter for the prototype radio frequency driven negative ion source for neutral beam injection.

In order to study the generation and extraction of negative ions for neutral beam injection application, a prototype radio frequency driven negative ion source and the corresponding test bench are under construction at the Institute of Plasma Physics, Chinese Academy of Sciences. A new design of infrared radiation diagnostic calorimeter for testing beam characteristics is put forward. Compared with the conventional calorimeter, the calorimeter adopts the block structure (8 × 28 tungsten hexahedron blocks) and modularization design (4 modules), so it has higher precision and good scalability. The thermal performance of the calorimeter is assessed using a finite element method. Simulation results show that the design can be achieved to operate in the stable-state mode at the maximum thermal flux 6.45 MW/m2 and meet the full requirement of beam diagnosis.

Thermal shock fracture behavior of wave-transparent brittle materials in hypersonic vehicles under high thermal flux by digital image correlation.

During diving, high orbit maneuver, or target detection and positioning, a hypersonic vehicle will experience high thermal flux. Aerodynamic heating will cause intense thermal shock to the antenna window and radar dome, which are usually made of brittle materials that are transparent to electromagnetic waves and are critical to the mission of target positioning and hitting. Therefore, determination of the time to fracture of the brittle materials under high thermal flux is of great significance. In this study, thermal shock tests were performed on two brittle materials, i.e., SiO2 and Al2O3, using a quartz lamp radiator test system with a maximum thermal flux of 1.5MW/m2. The difference between the pre-set and actual thermal fluxes for the tests was smaller than 1.0%. The time to fracture was determined for the two brittle materials by employing the digital image correlation method to capture and analyze changes in speckle images of the specimen's surface. The speckle image analysis also revealed variations in the surface strain values (εx and εy) prior to fracture. The test results provide important input for the safety and reliability design of radar domes and other electromagnetic wave-transparent signal detection and positioning components of hypersonic vehicles when subjected to high thermal fluxes.

Preliminary study for alternative conceptual core design of the MTR research reactor

The utilization of the research reactor is increasingly widespread, especially for radioisotope production and testing of advanced materials and preference to use a compact core. The reactor core design has been determined on the maximum thermal flux in the middle of the core per MW. BATAN has designed several alternative research reactor cores. The purpose of this research reactor is to obtain the optimum reactor core configurations with the criteria to have a thermal neutron flux in the centre of the core with minimum of 1.0×1015 n/cm2 s. Power level of the research reactor is 60 MWthwith U9Mo/Al fuel 85 cm of height. Design of plate-type fuels with a higher core results in the heat transfer to the coolant optimal. All 16 fuel assemblies and the 4 control rodsare inserted into the core for this reactor. The core design calculations were carried out with the WIMSD-5B and BATAN-FUEL codes. Conceptual design calculation results show that the core configuration with 5 × 5 grids, all the fresh fuel, fuel loading of 470 g, a D2O reflector, a maximum thermal neutron flux in the central core is 1.09 x 1015 n/cm2s and the cycle length is 33 days. The reactor core design is the most optimal for MTR type. For the equilibrium core, a fuel loading of 600 g results in the maximum thermal flux of 1.07x1015 n/cm2s and the two safety rods should be used in the core.

Investigation of some possible changes in Am-Be neutron source configuration in order to increase the thermal neutron flux using Monte Carlo code

Am-Be neutrons source is based on (α, n) reaction and generates neutrons in the energy range of 0-11 MeV. Since the thermal neutrons are widely used in different fields, in this work, we investigate how to improve the source configuration in order to increase the thermal flux. These suggested changes include a spherical moderator instead of common cylindrical geometry, a reflector layer and an appropriate materials selection in order to achieve the maximum thermal flux. All calculations were done by using MCNP1 Monte Carlo code. Our final results indicated that a spherical paraffin moderator, a layer of beryllium as a reflector can efficiently increase the thermal neutron flux of Am-Be source.

Flux trap effect study in a sub-critical neutron assembly using activation methods

ABSTRACTThe neutron flux trap effect was experimentally studied in the subcritical assembly of the Atomic and Nuclear Physics Laboratory of the Aristotle University of Thessaloniki, using delayed gamma neutron activation analysis. Measurements were taken within the natural uranium fuel grid, in vertical levels symmetrical to the Am–Be neutron source, before and after the removal of fuel elements, permitting likewise a basic study of the vertical flux profile. Three identical flux traps of diamond shape were created by removing four fuel rods for each one. Two (n, γ) reactions and one (n, p) threshold reaction were selected for thermal, epithermal and fast flux study. Results of thermal and epithermal flux obtained through the 197Au (n, γ) 198Au and 186W (n, γ) 187W reactions, with and without Cd covers, to differentiate between the two flux regions. The 58Ni (n, p) 58Co reaction was used for the fast flux determination. An interpolation technique based on local procedures was applied to fit the cross sections data and the neutron flux spectrum. End results show a maximum thermal flux increase of 105% at the source level, pointing to a high potential to increase in the available thermal flux for future experiments. The increase in thermal flux is not accompanied by a comparable decrease in epithermal or fast flux, since thermal flux gain is higher than epithermal and fast neutron flux loss. So, the neutron reflection is mainly responsible for the thermal neutron increase, contributing to 89% at the central axial position.

Evaluation of Main Parameters in Re-Entry Trajectories

The selection of the descending path for re-entry vehicles presents serious challenges for the designer. Basic requirements in terms of flight mechanics, dealing with initial and terminal conditions, have to match strict requirements related to the survivability of the re-entering body, in terms of structural integrity as well as in terms of thermal input and accepted temperature raise. The possibility of aero-braking assistance, allowed only if the targeted celestial body is surrounded by an atmosphere, increases the degrees of freedom available to the designer. The paper focuses on this special case, looking for an evaluation of the critical parameters (maximum load factor, maximum thermal flux) with respect to the flight dynamics of the trajectory. The initial step is represented by a procedure proposed by Broglio to approximately identify the initial conditions for suitable descent trajectories as function of a limited number of parameters. The solutions do not depend explicitly on the shape of the body, and the approximations involved are quite reasonable, as shown by a comparison with purely numerical integration. The present papers builds on this approach by modifying the analytical process to better handle the flight phases close to the deceleration peak, which are the most critical ones. Some examples are included to show the advantages of the proposed approach.

Dynamic modeling and motion precision analysis of spacecraft manipulator with harmonic drive considering the alternate thermal field in orbit

Nowadays, the harmonic drive is widely used as the reducer in the spacecraft manipulator, which may influence the dynamical properties of flexible spacecraft manipulator. The alternative thermal environment makes the spacecraft manipulator to experience periodic heating and cooling in the sunlight and shadow region of the Earth. The analysis of dynamic modeling and motion precision of flexible spacecraft manipulator with harmonic drive, considering the alternate thermal field in orbit is of significant importance for spacecraft manipulator designers in the early stage of design. The thermal load influences the motion precision, which reflects whether the mechanism is performed normally or not. In order to evaluate the loss of motion precision, this paper establishes the dynamical model of spacecraft manipulator with harmonic drive considering the alternate thermal field in orbit. A thermal analysis model of flexible spacecraft manipulator with harmonic drive is developed to characterize the thermal response of the whole spacecraft manipulator system subjected to space heat flux. Two different altitudes including low Earth orbit and geosynchronous Earth orbit are considered. Moreover, the transient temperature fields in different orbits of spacecraft manipulator and the effects of the thermal environment factors on the spacecraft manipulator are investigated. Simulation results reveal the evolution process of the transient temperature field of the spacecraft manipulator system. According to the results, the maximum temperature difference for space manipulator can lead to more severe precision loss compared with the minimum temperature difference. In addition, the vibration frequency of angular velocity error is determined by the maximum thermal heat flux. The proposed method is useful for forecasting the temperature distribution of the spacecraft manipulator system, and will provide meaningful information for performance enhancement of the aerospace facilities.

A core design concept for multi-purpose research reactors

A new core design concept for multi-purpose research reactors is presented. New concept cores using material test reactor (MTR) type fuel are constructed with the edge trimmed in-core irradiation hole. Conventional research reactors have core configurations with a constant assembly pitch. Sizes of in-core irradiation holes are limited to a multiple size of fuel assembly. The new concept enables a core configuration to have different sizes of in-core irradiation holes using the same size of fuel assembly. Two types of cores for multi-purpose research reactors are constructed and the performance results using MCNP calculation are compared. The new compact core for utilizing mainly beam tubes has a larger in-core irradiation hole and gives higher thermal neutron flux at the hole. The maximum thermal flux at the reflector region is almost the same. The new large core for utilizing in-core irradiation holes has small and large holes, while the conventional core has holes of the same size. The small hole is useful for high fast neutron flux and the large hole is useful for high thermal neutron flux. Compared to the conventional cores, the new cores use a smaller fuel assembly, which leads to enhanced safety at a high coolant speed for high power density. Two types of fuel plates are required for the conventional cores, but only one type of fuel plate will do for the new cores. This new design concept gives us several advantages and will be very useful for core design of multi-purpose research reactors.

Fully automated measurement setup for non-destructive characterization of thermoelectric materials near room temperature

A measurement setup is presented that allows for a complete and non-destructive material characterization of electrochemically deposited thermoelectric material. All electrical (Seebeck coefficient α, electrical conductivity σ), thermal (thermal conductivity λ), and thermoelectric (figure of merit ZT) material parameters are determined within a single measurement run. The setup is capable of characterizing individual electrochemically deposited Bi(2+x)Te(3-x) pillars of various size and thickness down to a few 10 μm, embedded in a polymer matrix with a maximum measurement area of 1 × 1 cm(2). The temperature range is limited to an application specific window near room temperature of 10 °C to 70 °C. A maximum thermal flux of 1 W/cm(2) can be applied to the device under test (DUT) by the Peltier element driven heat source and sink. The setup has a highly symmetric design and DUTs can be mounted and dismounted within few seconds. A novel in situ recalibration method for a simple, quick and more accurate calibration of all sensors has been developed. Thermal losses within the setup are analysed and are mathematically considered for each measurement. All random and systematic errors are encountered for by a MATLAB routine, calculating all the target parameters and their uncertainties. The setup provides a measurement accuracy of ±2.34 μV/K for α, ±810.16 S/m for σ, ±0.13 W/mK for λ, and ±0.0075 for ZT at a mean temperature of 42.5 °C for the specifically designed test samples with a pillar diameter of 696 μm and thickness of 134 μm, embedded in a polyethylene terephthalate polymer matrix.

Monte Carlo simulation of the thermal column and beam tube of the TRIGA Mark II research reactor

The Monet Carlo simulation of the TRIGA Mark II research reactor core has been performed employing the radiation transport computer code MCNP5. The model has been confirmed experimentally in the PhD research work at the Atominstitute (ATI) of the Vienna University of Technology. The MCNP model has been extended to complete biological shielding of the reactor including the thermal column, radiographic collimator and four beam tubes. This paper presents the MCNP simulated results in the thermal column and one of the beam tubes (beam tube A) of the reactor. To validate these theoretical results, thermal neutron flux density measurements using the gold foil activation method have been performed in the thermal column and beam tube A (BT-A). In the thermal column, the theoretical and experimental results are in fairly good agreement i.e. maximum thermal flux density in the centre decreases in radial direction. Further, it is also agreed that thermal flux densities in the lower part is greater than the upper part of the thermal column. In the BT-A experiment, the thermal flux density distribution is measured using gold foil. The experimental and theoretical diffusion lengths have been determined as 10.77 cm and 9.36 cm respectively with only 13% difference, reflecting good agreement between the experimental and simulated results. To save the computational cost and to incorporate the accurate and complete information of each individual Monte Carlo MC particle tracks, the surface source writing capability of MCNP has been utilized to the TRIGA shielding model. The variance reduction techniques have been applied to improve the statistics of the problem and to save computational efforts.

Optimization of core reload pattern for PARR-1 using evolutionary techniques

A procedure for searching optimum loading pattern to ensure safe and efficient reactor utilization was established for Pakistan Research Reactor-1 (PARR-1). A 6 × 5 core of PARR-1 consisting of 24 standard fuel elements, 5 control fuel elements and one flux trap was used in the study. The optimum loading schemes for maximum cycle length and maximum thermal flux in the flux trap were recommended using the developed methodology.

Fission MOLY production at PARR-1 using LEU plate type target

Neutronic analyses and depletion calculations have been performed for the production of 99Mo at PARR-1. Analysis has been performed with 20% enriched 235U target bearing plate type aluminized fuel (U–Al). Target (target holder and fuel plates) design contains three fuel plates and two aluminum dummy plates. Neutronic calculations were carried out for core at the beginning of equilibrium cycle of Pakistan Research Reactor-1 (PARR-1). Target analysis was performed by irradiating it at a location of maximum thermal flux available in the core. For this purpose, irradiation was performed at five different axial planes of the central water box facility. Thermal neutron flux profiles were also studied at different axial positions in available irradiation locations. Computer code WIMSD/4, a transport theory lattice code was employed for the generation of 10 group microscopic cross-sections. Diffusion theory code CITATION was utilized for three-dimensional modeling of the core. It was observed that from reactivity point of view, insertion or removal of target from the core will not affect the safety of reactor. Maximum heat flux in the target would be 102.68 W/cm 2 which is below the point of onset of nucleate boiling. However, forced flow is required to avoid initiation of nucleate boiling. The computer code ORIGEN2 was employed for depletion calculations. Analysis was performed for 100 Ci activity of 99Mo. After 100 days decay, waste activity will be less than 1 Ci and it will not pose any problem for handling radioactive waste.

An innovative research reactor design

A new and innovative core design for a research reactor is presented. It is shown that while using the standard, low enriched uranium as fuel, the maximum thermal flux per MW of power for the core design suggested and analyzed here is greater than those found in existing state of the art facilities without detrimentally affecting the other design specs. A design optimization is also carried out to achieve the following characteristics of a pool type research reactor of 10 MW power: high thermal neutron fluxes; sufficient space to locate facilities in the reflector; and an acceptable life cycle. In addition, the design is limited to standard fuel material of low enriched uranium. More specifically, the goal is to maximize the maximum thermal flux to power ratio in a moderate power reactor design maintaining, or even enhancing, other design aspects that are desired in a modern state of the art multi-purpose facility. The multi-purpose reactor design should allow most of the applications generally carried out in existing multi-purpose research reactors. Starting from the design of the German research reactor, FRM-II, which delivers high thermal neutron fluxes, an azimuthally asymmetric cylindrical core design with an inner and outer reflector, is developed. More specifically, one half of the annular core (0 < θ < π) is thicker than the other half. Two variations of the design are analyzed using MCNP, ORIGEN2 and MONTEBURNS codes. Both lead to a high thermal flux zone, a moderate thermal flux zone, and a low thermal flux zone in the outer reflector. Moreover, it is shown that the inner reflector is suitable for fast flux irradiation positions. The first design leads to a life cycle of 41 days and high, moderate and low (non-perturbed) thermal neutron fluxes of 4.2 × 10 14 n cm −2 s −1, 3.0 × 10 14 n cm −2 s −1, and 2.0 × 10 14 n cm −2 s −1, respectively. Heat deposition in the cladding, coolant and fuel material is also calculated to determine coolant flow rate, coolant outlet temperature and maximum fuel temperature under steady-state operating conditions. Finally, a more compact version of the asymmetric core is developed where a maximum (non-perturbed) thermal flux of 5.0 × 10 14 n cm −2 s −1 is achieved. The core life of this more compact version is estimated to be about 23 days.

Flux calculation in LSNAA using an 241Am-Be source

The CITATION code based on neutron diffusion theory is used for flux calculation inside voluminous sample in prompt gamma activation analysis with an isotopic neutron source ( 241 Am–Be). The code used the specific parameters related to energy spectrum source, irradiation system materials (shielding, reflector, etc.), geometry and elemental composition of the sample. The flux distribution (thermal and fast) was calculated on threedimensional geometry for the system: source, air, and polyethylene and water cylindrical sample of 125 liters. The thermal flux was calculated in series of points inside the sample, and agreed with the results obtained by measurements with good statistical uncertainty. The maximum thermal flux was measured at distance of 4.1 cm and calculated at 4.3 cm by the CITATION code. Beyond a depth of 7.2 cm, the ratio of thermal flux to fast flux increases up to twice and allows us the optimization of the detection system in the scope of in-situ PGNAA.