Keff Values Research Articles

Effect of the Peptidic Scaffold in Copper(II) Coordination and the Redox Properties of Short Histidine-Containing Peptides.

A linear decapeptide containing three His and one Asp residues and a β-turn-inducing dProPro unit was synthesised. A detailed potentiometric, mass spectrometric and spectroscopic study showed that at a 1:1 ratio of CCu /Cpeptide this peptide formed a major [CuH(O(dPro)-Asp)](2+) species (pH range 5.5-7.0), in which the Cu(2+) ion was bound to the His and Asp residues in square-planar or square-pyramidal geometries. The stability constant corrected for protonated species (log K* CuH(O dPro-Asp)=9.33) is almost equal to the value obtained for the parent [CuH(OAsp)](2+) species (log K*CuH(O-Asp) =9.28), but lower than that obtained for the cyclic [CuH(C-Asp)](2+) complex (log K*CuH(C-Asp) =10.79) previously published. Thus, the replacement of the ProGly unit by the stronger β-turn-inducing dProPro unit did not generate a more stable copper(II) species, although the O(dPro)-Asp peptide was structured in solution, as shown by circular dichroism (CD) spectroscopy. Interestingly, the calculated value of Keff showed that this peptide behaved similarly to the O-Asp or C-Asp counterparts, depending on the pH value. The cyclic voltammetry data indicated that the most easily reducible species were [CuH(O-Asp)](2+) (E'(0) =262 mV versus a normal hydrogen electrode (NHE)) and [CuH(O(dPro)-Asp)](2+) (E'(0) =294 mV versus NHE) complexes, the peptidic scaffolds of which are open. A lower value was obtained for [CuH(C-Asp)](2+) (E'(0) =24 mV versus NHE). A different degree of non-reversibility was observed for the three copper(II) complexes; this could reflect a different degree of flexibility in their respective peptidic scaffolds.

Weapons Grade Plutonium utilization with fertile materials in a Gas Turbine-Modular Helium Reactor

In this study, neutronic performances of the fertile thorium and natural uranium were studied in a Gas Turbine-Modular Helium Reactor using Weapons Grade Plutonium as fissile fuel. In the first step of the calculation procedure, a series of neutronic calculations were made in order to find the proper mixture ratios for these fertile fuel mixtures by using MCNP5 with ENDF/B-VI cross-section library. After the determination of these mixture ratios, in order to compare these fuel mixtures, time dependent neutronic calculations were made by using Monteburns 2.0 with MCNP5 and Origen 2.2. According to the results obtained, different parameters (operation time, burnup value, fissile isotope production, etc.) were compared. Results of the first step showed that, in order to reach the reference keff value, 690.6kg more fissile isotope was used in natural uranium mixture than the thorium mixture. In the second step of this study, time dependent calculations showed that for the natural uranium mixture, operation time and burnup values were found as 1935days and 177.76GWd/MTU, respectively, while these values for thorium mixture were 1925days and 153.95GWd/MTU.

Characterization of Piezoelectric Ceramic [Ba(Nd0.1Ti0.8Nb0.1)O3]0.40[Na0.5Bi0.5TiO3]0.40[CaTiO3]0.20

[BaNd0.1Ti0.8Nb0.1O3]0.40[Na0.5Bi0.5TiO3]0.40[CaTiO3]0.20 ceramic was prepared through solid state sintering method. The phase and microstructure of ceramic sample were characterized by X-ray diffraction (XRD) and Scanning electron microscopy (SEM). The XRD results indicate that the ceramic sample have a cubic structure at room temperature. The average grain size was found to be 2 4 μm. The dielectric and piezoelectric properties were also investigated. From capacitance measurements, the values of Kt, Keff and d33 at room temperature were 0.816, 0.788 and 15.4 pm.V−1 respectively. From modulus data, the Kt and Keff at 325°C are 0.9995 and 0.9994 respectively. The piezoelectric constant (d33) decreased with the increase of temperature and the value of d33 at 325 °C was found to be 8.58 nm.V−1.

Automatic monitoring of the effective thermal conductivity of snow in a low-Arctic shrub tundra

Abstract. The effective thermal conductivity of snow, keff, is a critical variable which determines the temperature gradient in the snowpack and heat exchanges between the ground and the atmosphere through the snow. Its accurate knowledge is therefore required to simulate snow metamorphism, the ground thermal regime, permafrost stability, nutrient recycling and vegetation growth. Yet, few data are available on the seasonal evolution of snow thermal conductivity in the Arctic. We have deployed heated needle probes on low-Arctic shrub tundra near Umiujaq, Quebec, (N56°34'; W76°29') and monitored automatically the evolution of keff for two consecutive winters, 2012–2013 and 2013–2014, at four heights in the snowpack. Shrubs are 20 cm high dwarf birch. Here, we develop an algorithm for the automatic determination of keff from the heating curves and obtain 404 keff values. We evaluate possible errors and biases associated with the use of the heated needles. The time evolution of keff is very different for both winters. This is explained by comparing the meteorological conditions in both winters, which induced different conditions for snow metamorphism. In particular, important melting events in the second year increased snow hardness, impeding subsequent densification and increase in thermal conductivity. We conclude that shrubs have very important impacts on snow physical evolution: (1) shrubs absorb light and facilitate snow melt under intense radiation; (2) the dense twig network of dwarf birch prevent snow compaction, and therefore keff increase; (3) the low density depth hoar that forms within shrubs collapsed in late winter, leaving a void that was not filled by snow.

Crack Propagation of 7050 Aluminum alloy Under Constant Amplitude Loading and Peak Overloads

Aluminium alloys of the 7000 series are largely used in aeronautical structural applications. In this work a fatigue crack propagation study was done in 7050 aluminium alloy under constant load amplitudes as well as under peak overloads. The tests were carried out using MT specimens in a servohydraulic machine at a frequency of 25Hz. For the constant amplitude tests three different stress ratios were analysed: R=0.05, 0.3 and 0.5. Crack closure was monitored in all tests by the compliance technique using a pin microgauge. The crack opening loads were derived from the load-crack opening displacement records after mathematics derivation and used to calculate □Keff values from which the da/dN-□□Keff curve is obtained. The crack closure parameter U was obtained and related with □K and the stress ratio. Under peak overloads the crack closure data show basically the same trend as the corresponding experimentally observed crack growth rate transient response.

Reactivity Prediction Calculations for Infinite Arrays of PWR MOX Fuel Pin Cells using MCNP5 and WIMSD-5B

In the present paper, the effect of some major nuclides on the reactivity of nuclear fuel for fuel pin cell in the Pressurised Water Reactor (PWR) was calculated using MCNP5_RSICC, 1.2 and WIMSD-5B codes. In order to assess our capabilities, the credit criticality Benchmark for infinite arrays of PWR MOX Fuel Pin Cells were calculated using two different codes and two different cross section libraries. Results were compared with similar published results by the Organization for Economic Co-operation and Development (OECD/NEA) Expert Group on Burn-up Credit. Good agreement was found for the value of keff with the corresponding codes and the reference mean solutions. The reactivity worth due to uranium, plutonium and curium isotopes was determined. The discrepancies between the present results and the reference mean results are within the range of the discrepancies for the participants with the reference mean results. The spread in the keff results with the mean results varies from -0.246 to 0.276 and from 0.05 to 0.793 for MCNP5 and WIMSD-5B respectively.

Magnetic Anisotropy of Maghemite Nanoparticles Probed by RF Transverse Susceptibility

We present radio frequency magnetic transverse susceptibility measurements on γ-Fe2O3 nanoparticles, which yield an estimation of their effective anisotropy constant, Keff as a function of nanoparticle size. The resulting values range from 4 to 8 × 104 erg/cm3, being on the order of the magnetocrystalline anisotropy in bulk maghemite. Keff values increase as the particle diameter increases. Evidences of anisotropy field distribution given by the size distribution in the samples, and interparticle interactions that increase as the particle size increases, are also observed in the TS measurements. The effects of such interparticle interaction overcome those of thermal fluctuations, in contrast with the behavior of other iron oxide particles.

Verification of the MCU precision code and ROSFOND neutron data in application to the calculations of criticality of fast reactors with highly enriched uranium

Calculation of 335 critical assemblies (benchmark experiments) with the core of highly enriched uranium and reflectors of various materials is performed. The statistical analysis of the results shows that, for all 16 materials studied, the absolute value of the most probable deviation of the calculated value of Keff from the experimental one does not exceed 0.005.

BRAHMMA: A compact experimental accelerator driven subcritical facility using D-T/D-D neutron source

A zero-power, sub-critical assembly BRAHMMA driven by a D-T/D-D neutron generator has been designed and commissioned at Bhabha Atomic Research Centre, India. This facility has been conceived for investigating the static and dynamic neutronics properties of accelerator driven sub-critical systems. This paper describes the design details of the system. Preliminary results of flux measurements and reactivity measurements using pulsed neutron source techniques have also been presented. This system has the advantage of being modular in design which enables its keff values to be changed depending upon type of fuel being used and it is first in the series of subcritical assemblies being designed.

Solution of the CB6 benchmark on VVER-440 final disposal using the Serpent reactor physics code

Abstract The CB6 benchmark on VVER-440 final disposal consists of two parts. The decay calculation part is aimed at determination of isotopic composition of irradiated fuel material over the time scale relevant to long-term spent fuel disposal. The second part involves calculation of keff values in a 3D cask configuration at various time points. In this study, the code Serpent is applied in solution of the benchmark. Serpent is a continuous energy Monte Carlo reactor physics code developed at VTT Technical Research Center of Finland. The decay calculation part of the benchmark is, in addition to Serpent, also carried out using point-depletion calculation code ORIGEN. This enables comparison of the results obtained using the two codes exploiting two different methods in solution of the decay equations. The paper demonstrates the decay and keff calculation results as determined in the benchmark specification. The differences in results arising between the two calculation codes in decay calculation, and between the ENDF/B-VII and JEFF-3.1.1 based cross section libraries in keff calculation, are depicted and discussed.

The Criticality Safety Analysis of the Spent Fuel Storage Pool Area I in Small Modular Reactor

Based on the Monte Carlo calculation method, geometric model of spent fuel storage pool Area I of small modular reactor is established, assuming infinite 6×6 type storage racks. Calculation results show that the reactivity is maximal when the water density is 1.0g/cm3. The value of keff is 0.8729 in normal storage condition. The spacing of storage racks in spent fuel pool would change in an earthquake accident condition. The values of reactivity of spent fuel pool in the assumed earthquake accident condition are also calculated. The values of keff are between 0.872 and 0.876. Both in normal condition and assumed earthquake accident condition, the values of keff are less than 0.95, to meet nuclear safety regulatory requirements.

Calculation of the NEA C5G7 MOX benchmark with MCNP5

As a part of the preparation of human building capabilities in Egypt in neutronic calculations for power reactors, a preliminary step was to perform calculations on benchmark problems related to neutronics calculations on lattice and core levels. Monte Carlo codes are capable of calculating integral parameters allowing all geometrical details of each individual fuel pin of each fuel assembly to be modeled. However, when more detailed information is required, it becomes much more difficult to get reliable results with respect to the associated statistical uncertainty within an acceptable computation time. Nonetheless, this is the requirement of current design calculations of nuclear reactor core. In order to assess the problems of Monte Carlo results for pin power distribution, this paper presents the results of the calculations of the C5G7 NEA benchmark problem using Monte Carlo code MCNP5. The present calculations used 50 & 90 million histories and 0.354% & 0.312% RMS statistical pin power percent error were achieved for the three-dimensional configurations. Results were compared with similar published results. Good agreement was found for the value of keff with the reference solutions. For the pin power distribution, discrepancy is large for UO2 assembly. The results for MOX assembly are much better.

Development of Monte Carlo-based pebble bed reactor fuel management code

A fuel management code for pebble bed reactors (PBRs) based on the Monte Carlo method has been developed in this study. The code, named Monte Carlo burnup analysis code for PBR (MCPBR), enables a simulation of the Once-Through-Then-Out (OTTO) cycle of a PBR from the running-in phase to the equilibrium condition. In MCPBR, a burnup calculation based on a continuous-energy Monte Carlo code, MVP-BURN, is coupled with an additional utility code to be able to simulate the OTTO cycle of PBR. MCPBR has several advantages in modeling PBRs, namely its Monte Carlo neutron transport modeling, its capability of explicitly modeling the double heterogeneity of the PBR core, and its ability to model different axial fuel speeds in the PBR core. Analysis at the equilibrium condition of the simplified PBR was used as the validation test of MCPBR. The calculation results of the code were compared with the results of diffusion-based fuel management PBR codes, namely the VSOP and PEBBED codes. Using JENDL-4.0 nuclide library, MCPBR gave a 4.15% and 3.32% lower keff value compared to VSOP and PEBBED, respectively. While using JENDL-3.3, MCPBR gave a 2.22% and 3.11% higher keff value compared to VSOP and PEBBED, respectively. The ability of MCPBR to analyze neutron transport in the top void of the PBR core and its effects was also confirmed.

Effect of Temperature on Segregation Coefficients of Impurities in Phosphorus

The present work is focused on the relationship between effective segregation coefficient keff and temperature of melting zone for purification of phosphorus by zone melting method. Values of keff at four temperatures of melting zone are obtained for zone pass n=1 at travel velocity of molten zone v=5×10−3 m·h−1 and initial impurity concentration C0≤10 μg·g−1. lnkeff is a linear function of 1/T. The keff values of Al, Ca, Cr, Fe, Cd and Sb increase with temperatures while that of Mg is almost constant. The purification is acceptable at lower temperature of melting zone such as 323 K. The variations of enthalpy and entropy between impurities and phosphorus in the liquid and solid phases are also presented.

Analysis on criticality properties and control strategies after reflooding of a damaged reactor core

In order to evaluate the crucial physical phenomena in a re-criticality of the reactor corium following a core reflood, a variety of criticality analyses have been performed with the design of a hypothetical core based on a severe accident condition including a core melt-down. This study aims to assess the critical mass of the corium within the core, to analyze the criticality change with the molten level and the coolant void reactivity (CVR), and to establish of practical strategies for a criticality control using H3BO3. The MCNPX 2.5.0 code was used to design the hypothetical accident core and to calculate the effective multiplication factor (keff), and the US Peach Bottom and Japan Fukushima nuclear power plant (NPP) was chosen for a reference model. The critical level was set to 0.95 for a conservative evaluation, and the consequential critical mass was indicated to be about 60kg only. The 548 fuel assemblies loaded in the core were assumed to be molten up to approximately 77%, the keff value therefore has a range of 1.03241±0.00194 to 1.40801±0.00157. The negative reactivity was increasingly inserted as the coolant density was decreased from 95% to 0%, and the keff value approaches the sub-critical state when the void fraction is above 30%. Finally, according to the boron injection into water, it was found that at least 600ppm 10B is required to have the sufficient criticality safety and to assure the sub-criticality condition following a severe core damage accident.

A parameter model of gas exchange for the seasonal sea ice zone

Abstract. Carbon budgets for the polar oceans require better constraint on air–sea gas exchange in the sea ice zone (SIZ). Here, we utilize advances in the theory of turbulence, mixing and air–sea flux in the ice–ocean boundary layer (IOBL) to formulate a simple model for gas exchange when the surface ocean is partially covered by sea ice. The gas transfer velocity (k) is related to shear-driven and convection-driven turbulence in the aqueous mass boundary layer, and to the mean-squared wave slope at the air–sea interface. We use the model to estimate k along the drift track of ice-tethered profilers (ITPs) in the Arctic. Individual estimates of daily-averaged k from ITP drifts ranged between 1.1 and 22 m d−1, and the fraction of open water (f) ranged from 0 to 0.83. Converted to area-weighted effective transfer velocities (keff), the minimum value of keff was 10−55 m d−1 near f = 0 with values exceeding keff = 5 m d−1 at f = 0.4. The model indicates that effects from shear and convection in the sea ice zone contribute an additional 40% to the magnitude of keff, beyond what would be predicted from an estimate of keff based solely upon a wind speed parameterization. Although the ultimate scaling relationship for gas exchange in the sea ice zone will require validation in laboratory and field studies, the basic parameter model described here demonstrates that it is feasible to formulate estimates of k based upon properties of the IOBL using data sources that presently exist.

Physical and Thermal Characterization of Red Mud Reinforced Epoxy Composites: An Experimental Investigation

The present paper deals with the effect of volume fraction of filler particles on the effective thermal conductivity (keff) of particulate filled polymer composites. A new class of epoxy based composites reinforced with an industrial waste i.e. red mud, with filler content ranging from 0 to 25 vol % has been prepared by conventional hand lay-up technique. keff of these fabricated samples are measured and then compared with the values obtained from themathematical model proposed by the authors in their earlier investigation and also with some existing models. A significantenhancement of about 135% in the value of keff is observedfor maximum filler loading. This comparison tells that,among all the models,the results obtained from the proposed model are in closest approximationwith the measured values. Some physical properties like density and void fraction together with morphological behavior of the entire fabricated specimens are also reported.

Experimental Investigation of Micro-sized Aluminium Oxide Reinforced Epoxy Composites for Microelectronic Applications

The present work aims at developing a class of particulate filled polymer composites consisting of epoxy as base matrix reinforced with micro-sized aluminium oxide (Al2O3) particles. A set of composites, with Al2O3 content ranging from 0 to 25 vol% have been fabricated by hand lay-up technique. Density (ρc), effective thermal conductivity (keff), glass transition temperature (Tg), coefficient of thermal expansion (CTE) and dielectric constant (ɛc) of these composites are measured experimentally. The measure values of keff are then compared with the values obtained from few established theoretical model and with the values calculated using the mathematical correlation developed by the authors. It is seen that the values obtained from the correlation developed by the authors fit well with measured conductivity values. Inclusion of Al2O3 in resin increases the keff and Tg whereas CTE of composite decreases favourably. The dielectric constant of the composite also found to get modified with filler content.

Monte Carlo calculations on transmutation of plutonium and minor actinides of pebble bed high temperature reactor

The paper aimed to maximize the fuel burnup performance of plutonium and minor actinides fueled pebble bed high temperature reactor (PBMR-400). The PBMR-400 was designed as a reference core. The neutronic calculations were performed by the code combination MCNP-ORIGEN-MONTEBURNS. In this study, neutronic performances of three different types of nuclear fuels (Reactor Grade Plutonium – RGPu; Weapon Grade Plutonium – WGPu and Minor Actinides – MAs) combined with natural uranium were conducted in a PBMR-400 full core. The neutronic performances were compared with the original uranium fuel designed for this reactor. Neutronic calculations showed that 9.6 wt % enriched uranium has a core effective multiplication factor (keff) of 1.2395. Corresponding to this keff values the natural UO2/RG-PuO2; natural UO2/WG-PuO2 and natural UO2/MAO2 mixture were found 70%/30%, 76%/24% and 63%/37%, respectively. The operation times were computed as ∼2000, ∼2500 and 1400 days whereas, the corresponding burnup values were obtained as ∼163 000, ∼194 000 and ∼116 000 MWD/T, respectively, for end of life keff set equal to 1.08.

Benchmark analysis of SPERT-IV reactor with Monte Carlo code MVP

The benchmark experiment of the SPERT-IV D-12/25 reactor core has been analyzed with the Monte Carlo code MVP using the cross-section libraries based on JENDL-3.3. The MVP simulation was performed for the clean and cold core. The estimated values of Keff at the experimental critical rod height and the core excess reactivity were within 5% with the experimental data. Thermal neutron flux profiles at different vertical and horizontal positions of the core were also estimated. Cadmium Ratio at different point of the core was also estimated. All estimated results have been compared with the experimental results. Generally good agreement has been found between experimentally determined and the calculated results.