Keff Values Research Articles

Effect of Recovery and Recrystallization on the Hall–Petch Relation Parameters in Submicrocrystalline Metals: I. Experimental Studies

Yield strength σy, macroelastic limit σ0, and effective grain-boundary hardening coefficient Keff in the Hall–Petch relation (\({\sigma _y} = {\sigma _0} + {K_{eff}}/\sqrt d \)) in the submicrocrystalline (SMC) materials produced by equalchannel angular pressing are experimentally studied. It is shown that, as compared to parameter σ0 and K in the Hall–Petch relation for coarse-grained metals, the SMC metals are characterized by higher values of σ0 and lower values of Keff. The critical grain size (d1) at which Keff in the σy–d–1/2 relations of SMC materials changes falls in the range 0.2–0.5 μm. The dependences of macroelastic limit σ0 and coefficient Keff on the annealing temperature are found to be determined by recrystallization. If abnormal grain growth develops in annealing of SMC metals, anomalous hardening is detected and a nonmonotonic temperature dependence of coefficient Keff takes place. In the case of conventional recrystallization at a high annealing temperature, SMC metals exhibit a smooth decrease in σ0 and an increase in Keff to the values of K characteristic of coarsegrained materials.

Cross-observables and cross-isotopes correlations in nuclear data from integral constraints

Most recent evaluated nuclear data files exhibit excellent integral performance, as shown by the very good agreement between experimental and calculated keff values over a wide range of benchmark integral experiments. However, the propagation of the uncertainties associated with those nuclear data to integral observables, generally produces calculated distribution which are much (3–5 times) wider than the experimental uncertainties. Reducing the variances of the evaluated data to achieve consistency at the integral level would lead to unreasonably narrow variances in the light of differential experimental data. One way of solving that paradox could be to allow, for different observables like fission cross-sections (σf), the prompt fission neutron spectra (χ), and the average multiplicity of fission neutrons ([see formula in PDF]) to be correlated in a Bayesian-like, Total Monte-Carlo approach, under constraints from integral experiments from the ICSBEP (International International Criticality Safety Benchmark Evaluation Project) benchmark compilation. Future developments will be highlighted and restrictions imposed by the current formatting of nuclear data will be discussed.

Comparative neutronic study of homogeneous and heterogeneous thorium fuel based core design in a lead-cooled fast reactor

In the present work the utilization of thorium is investigated in a lead-cooled fast reactor. Two thorium fuel strategies were analyzed: a homogeneous and a heterogeneous mixture of 232Th were set in the European Lead-cooled Fast Reactor core model. For this, the Monte Carlo Serpent-2.27 code and the JEFF-3.1 cross section library were used to perform the calculations. Three neutronic parameters were analyzed for each fuel strategy: the effective neutron multiplication factor (keff) for a cycle of 900 days, the Doppler constant and the reactivity effect of coolant density. With the homogeneous mixture of 232Th the neutron multiplication factor increases during the whole operating cycle (900 days), allowing a longer fuel cycle, keeping negative values for the Doppler constant and the reactivity effect of coolant density, with −692 pcm and −148 pcm, respectively. On the other hand, the addition of 232Th heterogeneously in the core showed that there is the necessity to have a reactivity excess around 2.2% Δk to fulfill the same fuel operating cycle. Even when there is a ThO2 blanket zone, the 233U production is not enough to increase the keff value and to enlarge the fuel operating cycle. Otherwise, the Doppler constant and the reactivity effect of coolant density are also negative values.

Potentiometry of the Reaction of Residual Lignin of Lignocellulose Powder Material with Chlorine Dioxide

The consumption rate of chlorine dioxide in the reaction with the residual lignin was studied using measurements of the potential of a redox couple ClO2/ClO 2 − in an aqueous suspension of lignocellulose powder material. The methodology was developed for the study of reaction kinetics of chlorine dioxide with the residual lignin using the first order reaction model with the initial lignin concentration in the reaction medium varying in the range of (1–17) × 10–4 М at the initial chlorine dioxide concentration of 4.48 × 10–4 М. The monochronic rate constant of the second order reaction of chlorine dioxide with the residual lignin was calculated by the dependence keff(I) = keff(II)[L]0 under the lignin excess conditions, it was 1462 ± 108 М–1 s–1. Considering the initial concentrations of reagents, it was found that, with increasing the degree of lignin conversion Θ from ≈0.27 to 0.54, its reactivity in the interaction with the chlorine dioxide is reduced, and the value of keff(II) decreased from ≈1280 to 900 М–1 s–1, which according to the theory of polychronic kinetics was a manifestation of the kinetic nonequivalence of the various ensembles of macromolecules of the residual lignin. Thus, it was experimentally shown for the first time that residual lignin of lignocellulose powder material had a kinetic nonequivalence in the interaction with chlorine dioxide.

Characterization of the interfacial structure and perpendicular magnetic anisotropy in CoFeB‐MgO structures with different buffer layers

In this study, we describe the deposition of Hf and Mo metal layers individually on Ta to compose new buffer layers, ie, Ta/Hf and Ta/Mo, where CoFeB/MgO stacks are deposited using magnetron sputtering. The synthesised Ta/Hf buffer has higher surface roughness, while the Ta/Mo buffer has lower surface roughness as compared with the Ta buffer. The surface roughness of the buffer appears to influence the interface of the subsequently deposited layers, resulting in rougher or smoother CoFeB/MgO interfaces. Additionally, we present a report on the magnetic properties of Ta, Ta/Hf, and Ta/Mo buffer samples. As the annealing temperature is below 200 °C, the saturation magnetisation (Ms) values for all buffer layers increase at similar rates, whereas the effective magnetic anisotropy energy (Keff) values increase at varying rates. After annealing at 350 °C, Keff reaches its maximum value for Ta/Hf and Ta/Mo buffer layers, whereas the CoFeB/MgO interface width decreases to a minimum value. The width increases as the annealing temperature is increased over 350 °C, and Keff gradually decreases with increase in the annealing temperature. The CoFeB/MgO interface width is primarily dependent on the buffer/CoFeB interface width, which is a critical parameter to obtain high perpendicular magnetic anisotropy (PMA) and high‐quality films. This work provides perspectives for understanding and controlling PMA from the viewpoint of interfacial structure.

Comparison of uranium plutonium nitride (U-Pu-N) and thorium nitride (Th-N) fuel for 500 MWth gas-cooled fast reactor (GFR) long life without refueling

Summary Comparison of uranium plutonium nitride and thorium nitride fuel for 500 MWth gas-cooled fast reactor has been done. Gas-cooled fast reactor is one type of generation IV reactor that can be operated in high temperature. Due to the high temperature, it can be used in hydrogen production. In this study, we compare the neutronic analysis of two fuel types, ie, uranium nitride fuel (U,Pu)N and thorium nitride fuel (Th,U233)N. The neutronic calculation uses SRAC2006 code system, and the data libraries use JENDL4.0. First, the fuel pin calculation (PIJ calculation) has been done to take the macro data that are used in CITATION calculation. Both uranium and thorium fuel use heterogeneous configuration with 3 variation fuel in the core. F1 is located in the central core, F2 middle core, and F3 outer core. The variation of fuel fraction is 40% until 65%, cladding 10%, and coolant 25% until 40%. For (U,Pu)N fuel, the diameter of the active core is 220 cm, and the height of the active core is 110 cm. And for (Th,U233)N, the diameter of the active core is 250 cm and the height of the active core is 150 cm. The reflector radial-axial width is 50 cm. For uranium plutonium nitride fuel, the type of the fuel in one core is varied; ie, F1 is 8%, F2 is 10%, and F3 is 12%. For thorium nitride fuel, the type of the fuel in one core also varied; ie, F1 is 7.8%, F2 is 8%, and F3 is 8.8%. The optimum value of thorium nitride is when fuel fraction of region F1 = 60%, F2 = 57.5%, F3 = 60%, the burn up time up to 20 years without refueling, max k-eff value is 1.0109, and max excess reactivity value is 1.08%. Neutronic analysis shows that both uranium and thorium fuel have excess reactivity value less than 2% but thorium fuel has excess reactivity less than uranium nitride fuel. Uranium fuel has better breeding capability than thorium fuel. Therefore, it is better to use uranium fuel for fast reactor like GFR, which has high breeding capability.

Processing and benchmarking of evaluated nuclear data file/b-viii.0β4 cross-section library by analysis of a series of critical experimental benchmark using the monte carlo code MCNP(X) and NJOY2016

Abstract To validate the new Evaluated Nuclear Data File (ENDF)/B-VIII.0β4 library, 31 different critical cores were selected and used for a benchmark test of the important parameter keff. The four utilized libraries are processed using Nuclear Data Processing Code (NJOY2016). The results obtained with the ENDF/B-VIII.0β4 library were compared against those calculated with ENDF/B-VI.8, ENDF/B-VII.0, and ENDF/B-VII.1 libraries using the Monte Carlo N-Particle (MCNP(X)) code. All the MCNP(X) calculations of keff values with these four libraries were compared with the experimentally measured results, which are available in the International Critically Safety Benchmark Evaluation Project. The obtained results are discussed and analyzed in this paper.

Study of the possibility of using Europium and Pyrex alloy as burnable absorber in PWR

The pursuit for an optimum burnable absorbers (BAs) takes a great interest due to their importance for regulating and control the reactivity of the reactor. MCNPX code is used to simulate the effect of Europium and Pyrex on the neutronic characteristics of PWR and compare it with the conventional burnable absorber. The thermal neutron flux and the power distribution through PWR assembly are analyzed to investigate the effect of gadolinium and europium on it. The PWR assembly is filled in a different fuel burnup process with 2% Gd2O3 and 2% Eu2O3 as integral burnable absorbers (IBAs). Despite the fact that the absorption cross section of gadolinium is larger than europium, the effective multiplication factor in the case of (UO2+Gd2O3) IBAs rods increases more than that in the case of (UO2+Gd2O3) IBAs rods. The Pyrex alloy is investigated in the Burnable poison rods (BPRs) to compare its effect on the neutronic characteristics of PWR with the conventional material such as B4C and (Ag-In-Cd). The effect of these materials on the Keff values has been analyzed. Due to the high absorption cross section of 135Xe and 149Sm resulting from the fission process, it was important to study the effect of different BAs materials on their concentration. The 239Pu and 241Pu concentrations resulting from the burning process have been calculated for the suggested BAs materials.

Scanning electrochemical microscopy investigation of the rate of formation of a passivating TiO2 layer on a Ti G4 dental implant

Titanium and alloys with titanium as the major component are widely used for making biomedical implants, such as artificial dental roots. In our laboratory, we have studied the kinetics of the self-healing reaction of the TiO2 film that forms on the surface of such an implant. Amperometric SECM approach curves were recorded over the surface of a grade 4 titanium (Ti G4) dental implant sample at specific times after the metal surface had been exposed to an air-saturated buffer solution. A ferrocene methanol redox mediator and a platinum microelectrode tip (r=12.5μm) were used in the experiments. The effective rate coefficient (keff) values for the mediator regenerating surface reaction were estimated using Wittstock's method from the approach curves recorded at different time points. Decreasing values of keff over time indicated an increasing rate of formation of the passivating TiO2 film.

Construction of FeN alloy films with ultra-strong magnetism and tunable magnetic anisotropy for spintronic application

FeN alloy film is a promising spintronic material with the theoretically ultra-strong magnetism (saturation magnetization MS and magnetic anisotropy Keff) and high spin polarization, which relies on the degree of N ordering interstice occupancy (S). However, due to the high activation energy for N ordering, the S value of an actual FeN film is mostly lower than 35% and this restricts the achievable magnetism and transportation property. Thus, the construction of a FeN alloy film with well-controlled magnetism and efficient electronic transportation remains a long-standing challenge. Here, we tackle the problem by strain engineering. Using an Fe/Cr underlayer, we introduced a considerable epitaxial strain in the FeN lattice. The strain is proven to effectively promote the S value to over 60%, resulting in remarkable enhancement of MS value from 2.18T to 2.81T (30% increment) and effective tunability of Keff value ranging 1.3∼2.2 × 106 J/m3. Besides, the matched energy band symmetry (Δ5) between Cr and Fe16N2 facilitates the efficient electronic transportation for spintronic applications. By simulating interstice distribution with the first-principles calculations, the lattice strain is found to decrease the activation energy for N interstitial migration, which serves as a thermodynamic driving force for the magnetism tunability.

Dielectric relaxor, impedance relaxor, PTCR and electromechanical effects in multifunctional ceramic: [Ba(Nd0.1Ti0.8Nb0.1)O3]0.65[Na0.5Bi0.5TiO3]0.25[BaZrO3]0.10

ABSTRACT[BaNd0.1Ti0.8Nb0.1O3]0.65[Na0.5Bi0.5TiO3]0.25[BaZrO3]0.10 (BNNT-NBT-BZ) ceramic is prepared through solid state sintering method. The dielectric, impedance and DC resistivity studies are carried out as a function of temperature. Dielectric and impedance studies revealed relaxor behaviour. DC resistivity studies revealed PTCR effect in the present ceramic. Capacitance resonance and anti-resonance studies in free and clamped state are carried out over a wide range of frequency at room temperature. The values of Kt, Keff and d33 at room temperature are 0.112, 0.105 and 1.003 pm.V−1 respectively for ceramic sample studied under free state, whereas for the samples studied under clamped state the values are 0.667, 0.629 and 26.410 pm.V−1 respectively. Stress and temperature induced modulus resonance and anti-resonance studies are carried out over a wide frequency range and temperature. The values of Kt, Keff and d33 at 723K are 0.99926, 0.99911 and 14.188 nm.V−1 respectively.

Comparative Study on Various Geometrical Core Design of 300 MWth Gas Cooled Fast Reactor with UN-PuN Fuel Longlife without Refuelling

Nuclear power has progressive improvement in the operating performance of exiting reactors and ensuring economic competitiveness of nuclear electricity around the world. The GFR use gas coolant and fast neutron spectrum. This research use helium coolant which has low neutron moderation, chemical inert and single phase. Comparative study on various geometrical core design for modular GFR with UN-PuN fuel long life without refuelling has been done. The calculation use SRAC2006 code both PIJ calculation and CITATION calculation. The data libraries use JENDL 4.0. The variation of fuel fraction is 40% until 65%. In this research, we varied the geometry of core reactor to find the optimum geometry design. The variation of the geometry design is balance cylinder; it means that the diameter active core (D) same with height active core (H). Second, pancake cylinder (D>H) and third, tall cylinder (D<H). The reflector radial-axial width is 50 cm. The power is 300 MWth. First calculation, we calculate survey parameter for UN-PuN fuel with fissile contain from Plutonium waste LWR for each geometry. The minimum power density is around 72 Watt/cc, and maximum power density 114 Watt/cc. After we calculate with various geometry core, when we use the balance geometry, the k-eff value flattest and more stable than the others.

The influence of natural and synthetic fibers on low temperature mixed mode I + II fracture behavior of warm mix asphalt (WMA) materials

In this contribution, low temperature crack growth resistance of Warm Mix Asphalt (WMA) mixtures reinforced with two types of fiber, namely natural Jute fiber and synthetic Polyolefin-Aramid (FORTA) fiber is investigated experimentally. Using semi-circular bend (SCB) specimen, the influence of different fiber weight percentages (0.3%, 0.5% and 0.7%) and test temperatures (0°, −10° and −20°C) on mixed mode I+II fracture toughness values (i.e. KIc and KIIc) are studied. It was shown that both fibers can increase in general the fracture resistance of WMA concrete relative to the control mixture. Moreover, modes I and II fracture toughness and effective fracture toughness (Keff) values are increased by decreasing the test temperature or increasing the fiber content. The FORTA fiber provided better crack growth resistance characteristics than the Jute fiber for the WMA mixtures. The effects of both natural and synthetic fibers on enhancing the performance of the tested WMA mixtures was more pronounced for dominantly pure mode I loading case.

Influence of reflector materials and core coolant on the characteristics of accelerator driven systems

In this paper we simulated the behavior of a simple ADS model, based on MOX fuel embedded in solid lead, in terms of multiplication coefficient keff, thermal power and absolute neutron spectra. Our focus on a subcritical, accelerator-driven system is motivated by the need to guarantee enhanced safety based on the distance from criticality and the dependance of the chain reaction on the external neutron source provided by the accelerator. The studies presented here are meant to expand the scientific reach of a previously proposed low-power ADS as the core apparatus of a research and training center in nuclear technologies, such as to increase the flexibility of the system in terms of accessible neutron spectra. In the first part of the paper, we report on the results obtained when modifying the reflector surrounding the fission core, by replacing pure lead with a layered graphite/lead structure. We found that, by appropriately choosing position and thickness of the graphite and lead layers, it is possible to obtain a “hybrid” system where the neutron spectrum in the core still exhibits a fast character, while the spectrum in the graphite layer is considerably softer, becoming thermal in the most peripheral positions. In order to obtain such a modulation of the neutron spectra from the center of the system to the periphery, a careful choice of the materials has to be made in order to avoid large variations of the local power at the core boundary. However, the smoothness of the power distribution is obtained at the expense of lower values of keff and the total power of the system. In the second part of the paper, we explored the option of adopting light water as coolant, instead of the helium gas assumed in the initial design. We found that this produces an increase in keff and thermal power, without significantly perturbing the fast character of the system and without introducing spatial power excursions in any place within the core. The characteristics obtained may allow to design a system where fast, mixed and thermal spectra can be used to expand the scope of the ADS as an irradiation facility.

Selection of Conservative Assumptions in Nuclear Safety Justification of SNF Storage Systems

The paper addresses the description of computer model for the spent fuel assemblies storage system in SCALE and MCNP codes, as well as the results in selection of conservative assumptions made to justify the nuclear safety of moving, transport and storage operations with the VVER-1000 spent nuclear fuel (SNF) in designed Centralized Spent Fuel Storage Facility (CSFCF). When justifying the nuclear safety, it is necessary to confirm that the maximum value of the effective multiplication coefficient K eff in SNF storage systems is kept below specified limit of 0.95 in any design-basis operation mode. The paper presents calculation results and analysis on the selection of the most conservative conditions of neutron multiplication leading to the maximum value of Keff.

Preliminary Study on Small Long-Life PWR Reactor Design Fueled Thorium with Protactinium as Burnable Poiso

Thorium as fissile material has overweigh advantages compared to Uranium in order to have a long-life PWR reactor without replacement fuel. Additional of Protactinium as a burnable poison is effective to reduce excess reactivity in the beginning of the reactor operation. This study aims to analyze the neutronic, burnup fuel, thermal-hydraulic and kinetic calculations in regard to reactor safety. The calculation is performed with SRAC program using JENDL-32 as data library. The results show that the value of keff for 33 years are still in the range 1.0154 &lt;keff&lt;1.0171 and the burn-up level has increased, due to the additional fissile material as conversion results of Th232 and Pa231. Axial linear power distribution indicates normal distribution pattern. The coolant temperature distribution in the axial direction is quite flat. Key words: PWR, Th-232, U-233, Pa-231, neutronic, thermal-hydraulic, burn-up level, reactor kinetics

Inverse kinetics for subcritical systems with external neutron source

Nuclear reactor reactivity is one of the most important properties since it is directly related to the reactor control during the power operation. This reactivity is influenced by the neutron behavior in the reactor core. The time-dependent neutrons behavior in response to any change in material composition is important for the reactor operation safety. Transient changes may occur during the reactor startup or shutdown and due to accidental disturbances of the reactor operation. Therefore, it is very important to predict the time-dependent neutron behavior population induced by changes in neutron multiplication.Reactivity determination in subcritical systems driven by an external neutron source can be obtained through the solution of the inverse kinetics equation for subcritical nuclear reactors. The main purpose of this paper is to find the solution of the inverse kinetics equation the main purpose of this paper is to device the inverse kinetics equations for subcritical systems based in a previous paper published by the authors (Gonçalves et al., 2015) and by (Gandini and Salvatores, 2002; Dulla et al., 2006). The solutions of those equations were also obtained. Formulations presented in this paper were tested for seven different values of keff with external neutrons source constant in time and for a powers ratio varying exponentially over time.

Separation of Cs and Sr from LiCl-KCl eutectic salt via a zone-refining process for pyroprocessing waste salt minimization

The purification of a LiCl-KCl salt mixture was carried out by a zone-refining process. To improve the throughput of zone refining, three heaters were installed in the zone refiner. The zone-refining method was used to grow pure LiCl-KCl salt ingots from a LiCl-KCl-CsCl-SrCl2 salt mixture. The main investigated parameters were the heater speed and the number of passes. From each zone-refined salt ingot, samples were collected axially along the salt ingot and the concentrations of Sr and Cs were determined. Experimental results show that the Sr and Cs concentrations at the initial region of the ingot were low and increased to a maximum at the final freezing region of the salt ingot. Concentration results of the zone-refined salt were compared with theoretical results furnished by the proposed model to validate its predictions. The keff values for Sr and Cs were 0.55 and 0.47, respectively. The correlation between the salt composition and separation behavior was also investigated. The keff values of the Sr in LiCl-KCl-SrCl2 and the Cs in LiCl-KCl-CsCl were found to be 0.53 and 0.44, respectively, by fitting the experimental data into the proposed model.

Development and verification of a new nuclear data processing system FRENDY

ABSTRACTJapan Atomic Energy Agency developed the new nuclear data processing system FRom Evaluated Nuclear Data librarY to any application (FRENDY) in order to solve the problems of the current widely used nuclear data processing systems and process the new evaluated nuclear data file. Verification of FRENDY was carried out by three steps, i.e. verification of each function, comparison of the results, and comparison of the keff values for the 79 benchmark experiments in the ICSBEP handbook using cross section data library processed by FRENDY with those by NJOY99. These results verified that FRENDY generates the ACE (A Compact ENDF) file correctly.

Neutronic Analysis of Thorium Nitride (Th, U233)N Fuel for 500MWth Gas Cooled Fast Reactor (GFR) Long Life without Refueling

Neutronic analysis of Thorium Nitride (Th, U233)N fuel of 500MWth Gas Cooled Fast Reactor (GFR) has been done. In this study the neutronic analysis use SRAC2006 code both PIJ and CITATION calculation. The data libraries use JENDL 4.0. First calculation is survey parameter with U-233 enrichment variation. From the homogeneous core configuration calculation, when the enrichment of U-233 is 8.2%, the maximum k-eff value is 1,00819 with excess reactivity value 0,812%. The average power density is 63 Watt/cc and the maximum power density 100 Watt/cc. The heterogeneous core configuration calculation has been done to flattening the power of the reactor. The variation fuel of F1:F2:F3 = 7.8%:8%:8.8%. The fraction of fuel : cladding: coolant = 60%:10%:30%. The max k-eff value of heterogeneous core configuration is 1,01229 with excess reactivity value 1.21%. The average power density is 65 Watt/cc and the maximum power density 92 Watt/cc. The power density distribution of heterogeneous core configuration is flatter than homogeneous core configuration.