Keff Values Research Articles

Enhancement of perpendicular magnetic anisotropy in W/Co/Pt films by nitrogen doping in the W layer

The modulation of perpendicular magnetic anisotropy (PMA) in films has been the subject of considerable research interest, as it is proposed to be a key component for the design and realization of efficient magnetic switching in spintronic devices. In this study, we report the appearance of PMA in the as-deposited WNx/Co/Pt films without annealing. The strength of the PMA is quantified by means of effective magnetic anisotropy constant Keff, which is correlated with the N2 gas/Ar gas flow rate ratio PN2. The highest Keff value, 1.347 × 106 erg/cm3, is obtained for the sample deposited with PN2 of 40%. This phenomenon can be explained in two ways. On the one hand, the results of the experiment demonstrate that appropriate nitrogen doping can facilitate the formation of an ideal nitrided state at the WNx/Co interface, while simultaneously reducing the roughness of the WNx/Co interface, which, in turn, enhances the PMA of the WNx/Co/Pt films. On the other hand, the first-principles calculations indicate that the enhancement of PMA can be attributed to the modification of orbital hybridization at the Co/Pt interface by WNx. This innovative approach has the potential to advance the development of high-performance magnetic random-access memory devices.

The Effect of Adding Minor Actinide Fuel Rods on GFR Reactor in Radiopharmaceutical Waste Production Using OpenMC Program

GFR is a generation IV reactor based on helium gas refrigeration capable of working at very high temperatures. The fast spectrum in this reactor makes it possible to use nitride-based fuel, namely Uranium Plutonium Nitride (UN-PuN). Adding minor actinide (MA) material to the primary fuel, UN-PuN can maximize reactor performance to near critical from the beginning to the end of burn-up. This study aims to analyze the effect of adding MA fuel rods to the heterogeneous core of 5 fuel variations (F1, F2, F3, F4, F5) on the probability of radiopharmaceutical waste production. The method in this research is to place MA fuel rods in this study using four designs based on the highest neutron flux value in one fuel assembly. The results of the neutron flux calculation show that the reactor’s active core’s central region (F1, F2, F3) needs to be added to MA fuel rods so that the resulting flux is more evenly distributed. The calculation of reactor criticality shows that Np fuel rod design 4 and Am fuel rod design 1 have the best keff value (keff ≈ 1) among other designs. The burn-up of MA fuel rods produces a minimal probability of producing Tc99m, Sr89, Y90, Rh105, Ag111, I231, and Sm15 radiopharmaceutical waste, even less than 1 kg.

Comparison of Neutronic Analysis for 250 MWth Molten Salt Reactor using Monte Carlo OpenMC Code with Different Nuclear Data Libraries of JENDL 5.0, ENDF/B-VII.1, ENDF/B-VIII.0, and JEFF 3.3

Abstract Nuclear data libraries playian important role in the accuracy of neutronic aspect calculations, which determine various factors in a nuclear reactor design. Likewise, deterministic or Monte Carlo methods significantly affect the calculation results. Thus, the combination of nuclear data library and calculation methods needs to be studied carefully. This research will conduct neutronic analysis for the Molten Salt Reactor (MSR) design, one of the Generation IV reactor types with advanced features for future energy resources, using the Monte Carlo-based program OpenMC and different nuclear data libraries. MSR will be designed for a 250 MWth power capacity, operable for five years. The fuel used will be LiF-BeF2 as a coolant and ThF4-UF4 as the fuel. The compositions for each fuel will be optimized to achieve critical conditions during five years of operation. This research aims to evaluate the use of differentinuclear data libraries on reactor criticality conditions, using the nuclear data libraries ENDF/VIII-B, ENDF/VII.0, JEFF 3.3, and JENDL 5.0. To perform the calculation and analysis, MSR chain depletion modification has to be done for different nuclear data libraries. Then, evaluation will be conducted on effective multiplicationifactor (k-eff), neutron flux spectrum, and cross-section. The results of this research show that the use of JENDL 5.0 provides more optimal critical conditions. This is due to the completeness of JENDL 5.0 nuclear data covering 795 nuclide data and higher fission cross-section values. While ENDF/VIII-B has 557 nuclide data evaluations compared to ENDF/VII.0’s 423 nuclide data, more stable k-eff values are shown for the use of ENDF nuclear data libraries. This research is expected to significantly contribute to meeting Indonesia’s energy needs by providing efficient solutions in MSR design in neutronic aspects using the most accurate nuclear data library.

Optimization of Fuel Enrichment and Reactor Core Pitch Dimension in Hydride Microreactor using NSGA-II Method and OpenMC Calculation

Abstract This study focuses on the design optimization of a hydride microreactor, aimed at supporting the equitable development of electricity in remote regions of Indonesia. Its compact design, inherent safety properties, and independence from the conventional electric grid make this microreactor concept suitable for electrifying even the most inaccessible areas. Current research investigates various aspects of the hydride microreactor design, employing a standard reactor core pitch dimension of 2.4 cm and fuel enrichment of 12%. It is crucial to operate the reactor using specific concentrations of Gadolinia burnable poison to enable a shutdown in the event of a stuck rod. By optimizing only the pitch dimension and fuel enrichment, the reactor can be simulated relatively quickly without the need for burnup calculations. Ideal configuration aims for lower enrichment, an extended fuel cycle, and a relatively low power peaking factor (PPF). The optimization process utilizes the Non-dominated Sorting Genetic Algorithm – II, with the neutron multiplication factor (keff) and PPF as primary objectives. Fuel cycle and feedback reactivity are the sub-objectives that are used to identify the optimal solution from the Pareto front. The calculation process of keff, PPF, fuel cycle, and feedback reactivity coefficients for specific pitch dimensions and fuel enrichment are done by utilizing the OpenMC Monte Carlo code. The optimal configuration is obtained through four generations of NSGA-II, featuring a reactor core pitch of 2.1 cm and fuel enrichment of 10.4%. This configuration yields a keff value of 1.0712049 and a PPF value of 2,0728 at the Beginning of Life (BOL). Remarkably, this configuration enables safe operation without using burnable poison for a continuous period of 12 years, generating 1 MW of thermal power. Emphasizing the reduction of fissile material is also essential to enhance the inherent safety of the reactor.

Chain-Folded Lamellar Stacking Structure of the Crystalline Fraction of Bombyx mori Silk Fibroin with Silk II Form Studied by 2D 13C-13C Homonuclear Correlation NMR Spectroscopy.

The structure of Bombyx mori silk fibroin (SF) is a subject of significant interest due to its remarkable physical properties; however, its atomic-level structure is still not conclusive. We previously proposed a lamellar stacking structure for the crystalline fraction (Cp) with β-turns occurring every eighth amino acid. In this study, we took the following steps: At first, a model of the chain-folded lamellar stacking structure in antipolar and antiparallel β-sheet layers was constructed. Then, dipolar-assisted rotational resonance solid-state NMR spectra were observed to determine the effective internuclear distance (rj,keff) for the uniformly 13C-labeled Cp fraction sample. By comparing the experimentally obtained rj,keff (obs) values with the calculated rj,keff (calc) values from our structural model, a fairly good correlation between the observed and calculated values of the internuclear distances was obtained with a standard deviation of 0.37 Å. This supports the existence of the chain-folded lamellar stacking structure in the SF fiber. These findings contribute to our understanding of the atomic-level structure of SF and its exceptional properties.

Acoustic and thermal performance of wood strands-rock wool-cement composite boards as eco-friendly construction materials

With a growing emphasis on eco-friendly building materials, wood strands-rock wool-cement boards (WRCB) offer a promising solution due to their composition of renewable wood fibers and cementitious binder. This study delves into the intricate balance between sound absorption, thermal insulation, and cost-effectiveness of WRCB, all of which are crucial factors shaping indoor comfort, energy efficiency, and the economic viability of constructions. WRCB incorporating wood strands, Portland cement, rock wool, and calcium chloride were manufactured at various thicknesses (20–60 mm) and densities (400, 500, and 600 kg/m³). The WRCB exhibited sound absorption average (SAA) and effective thermal conductivity (Keff) values ranging from 0.28 to 0.56 and 0.285–0.381 W/(mK), respectively. These results underscore the remarkable SAA and Keff of the panels, showcasing their potential as sustainable construction materials. Notably, WRCB with thicknesses ranging from 30 mm to 50 mm and densities of 400–500 kg/m³ exhibited nearly ideal absorption levels between 1000 and 2000 Hz, indicating their practical suitability for speech-related scenarios in architectural environments. The utilization of the Response Surface Methodology allowed the optimal conditions for maximizing SAA while simultaneously minimizing Keff and cost to be identified through the optimization process. A density of 452.8 kg/m³ and a thickness of 37.8 mm were determined as the optimal parameters. Under these conditions, the resultant SAA reached 0.554, with a corresponding cost of 1724 IRR and Keff of 0.317 W/(mK). The results also demonstrated the significant impact of thickness and density on the acoustic and thermal performance of the WRCB. The acoustic performance of WRCB was further analyzed through predictive modeling using the Johnson-Champoux-Allard (JCA) and Delany-Bazley (DB) models. The results revealed that the JCA model exhibited superior predictive accuracy compared to the DB model.

Experiments on Criticality and Reactivity Worths in the FCA-XXII-1 Assembly Simulating Highly Enriched MOX Fueled Tight Lattice LWR Cores

A series of integral experiments were conducted at the fast critical assembly (FCA) of the Japan Atomic Energy Agency, simulating light water reactor cores with a tight lattice cell of highly enriched mixed-oxide (MOX) fuel containing >15% fissile plutonium (Pu). The three experimental configurations of the FCA-XXII-1 assembly were constructed using foamed polystyrene with different void fractions (45%, 65%, and 95%) to clarify the prediction accuracy of neutronics calculation codes and nuclear data libraries among various neutron spectra. The hydrogen-to–nuclear fuel atomic ratio varied from 0.1 to 0.8. The nuclear characteristics measured in the experiments were criticality (keff), moderator void reactivity worths, and sample reactivity worths using boron carbide (20%, 60%, and 90% 10B enrichment) and Pu (92%, 81%, and 75% fissile Pu ratio). Preliminary analyses on experiments were conducted using a deterministic calculation code system conventionally used for fast reactors and the Japanese evaluated nuclear data library of JENDL-4.0. The calculated keff values overestimated the experiments beyond the experimental uncertainties. However, most reactivity worth calculations agreed well with the experimental values. Even beyond the experimental uncertainties, discrepancies between the calculation and the experiment were <13%. Specifically in the reactivity worth analyses of the softer neutron spectra configurations, the treatment of ultrafine energy groups obviously improved the prediction accuracy of the deterministic calculations. Furthermore, reference calculations for criticality and large reactivity worths were performed with the Monte Carlo calculation code MVP3 by modeling the experimental configurations in detail, confirming that the deterministic calculations closely agreed with the reference values.

A neural based modeling approach for predicting effective thermal conductivity of brewer’s spent grain

PurposeBrewer's spent grain (BSG) is the main by-product of the brewing industry, holding significant potential for biomass applications. The purpose of this paper was to determine the effective thermal conductivity (keff) of BSG and to develop an Artificial Neural Network (ANN) to predict keff, since this property is fundamental in the design and optimization of the thermochemical conversion processes toward the feasibility of bioenergy production.Design/methodology/approachThe experimental determination of keff as a function of BSG particle diameter and heating rate was performed using the line heat source method. The resulting values were used as a database for training the ANN and testing five multiple linear regression models to predict keff under different conditions.FindingsExperimental values of keff were in the range of 0.090–0.127 W m−1 K−1, typical for biomasses. The results showed that the reduction of the BSG particle diameter increases keff, and that the increase in the heating rate does not statistically affect this property. The developed neural model presented superior performance to the multiple linear regression models, accurately predicting the experimental values and new patterns not addressed in the training procedure.Originality/valueThe empirical correlations and the developed ANN can be utilized in future work. This research conducted a discussion on the practical implications of the results for biomass valorization. This subject is very scarce in the literature, and no studies related to keff of BSG were found.

Generalized extreme value analysis of efficient evaluation of extreme values in random media criticality calculations

The uncertainty of fuel debris criticality can be estimated by Monte Carlo criticality calculations repeated over independent replicas of some properly modelled random medium. Incomplete randomized Weierstrass function (IRWF) is a modelling framework for such calculations. Under these approaches at hand, the theme of this paper is how to efficiently analyse extreme realizations of neutron effective multiplication factor (keff) over IRWF random media replicas. To this end, a new bounded amplification (BA) technique is applied to IRWF. The numerical results clearly indicate that the BA-applied IRWF reduces a required number of random media replicas at least by an order of magnitude. To rigorously validate this efficiency gain, generalized extreme value (GEV) analysis is applied to a data set of keff values obtained without applying BA. It turns out that the extreme values of these keff values follow the Weibull distribution. Therefore, the theory of GEV guarantees the existence of an upper limit for these keff values, and the actually computed upper limit is indeed smaller than or equal to the top two keff values obtained from an order-of-magnitude reduced number of BA-applied IRWF random media replicas. This means that the efficiency gain brought by BA has been confirmed by the GEV methodology. In addition, a method of rejecting spurious estimation values due to statistical fluctuation is demonstrated concerning the extreme value index in the GEV analysis.

Neutronic design analysis of CEA Cadarache URANIE subcritical assembly

This paper aims to study neutronic modeling using decommissioned fuel pins of the URANIE subcritical assembly at the Cadarache center of CEA for the future Tunisian Subcritical Assembly. Indeed, the CNSTN intends to exploit the fuel to implementing its subcritical assembly. Therefore, a comprehensive computational study of the URANIE fuel was carried out to analyze the neutronic parameters for future facilities including criticality safety, the neutron flux, and neutron spectrum, for future facilities. Indeed, MCNP Monte Carlo code was used to develop a three-dimensional sketch for URANIE core. Two versions were used to estimate an effective multiplication factor (keff): MCNP5 and MCNP6. The keff was varied from 0.85842 to 0.86463 with a maximal difference of 623 pcm and depended on ENDF/B data library releases. The nominal value of keff is acquired at the optimum lattice pitch of 4.6 cm, 30 cm in reflector thickness, and 185 cm in moderator level. Moreover, the reactivity of the facility is always negative. In addition, the energy spectrum was depicted at the three planes using 283 energy groups. Furthermore, the radial neutron flux was evaluated at eight positions. The axial neutron flux was estimated at various spatial planes of the assembly core. Inside the reactor core, the findings revealed indicate that the fast flux is higher and can be reach 7 × 10−4 n/cm2.Q and decreased steadily moving away from the center to the reflector part about 8 × 10−4 n/cm2.Q.

Demonstrating a Quantitative and Systematic Approach to Reducing Excess Conservativism in Nuclear Criticality Safety Analyses

Although reducing conservatism would alleviate unnecessary constraints in processing, storage, transportation, and disposal of nuclear materials, excessively conservative approaches are still utilized in many safety analyses. Criticality safety limits are put in place to reduce the likelihood of having a nuclear criticality accident to a value that is deemed incredible but often utilize parameters that are conservative to the point of becoming incredible themselves. The analyses that determine criticality limits are supposed to be based on credible instead of incredible events and circumstance, highlighting the need to be able to distinguish between what is in the realm of possibility and what is not. This paper provides a quantitative approach for reducing unrealistically conservative parameters by recalculating limiting factors in a state that deviates from the worst-case scenario and assigning probability distributions to these systematic deviations. This provides a technical basis for replacing excessively conservative values with something that is both objective and reasonably bounding, which may be systematically utilized in any criticality safety analyses. The assumption of “perfect sphericity” in the TRUPACT-II package’s fissile contents model was used as an example case to demonstrate the proposed approach for replacing qualitative reductions to conservatism with quantitative reductions. Through a series of Monte Carlo calculations and statistical analyses, it was shown that conservative deviations from sphericity will provide lower keff values, where the magnitude and impact of this deviation is system specific. The statistical significance from applying probabilistic conservatism will be dependent on the chosen κ value and integration limit for the exponential distribution, as it varies the degree of conservatism applied to any parameter of interest. This approach is not limited to geometric assumptions and may be applied to a variety of conservative parameters. In an effort to move toward a standard method for reducing conservatism, this objective approach may be used in lieu of or in conjunction with subjective methods for relaxing constraints.

Thickness dependence of structural and magnetic properties of electrodeposited Co2FeSn films

Co2FeSn films have been electrochemically deposited on polycrystalline Cu substrates in potentiostatic mode. Processed films with thicknesses of 320 nm, 780 nm, and 1400 nm exhibited highly ordered L21 type crystal structure with saturation magnetizations of 1023.38 kA/m (5.08 ± 0.04 μB/f.u.), 1035.31 kA/m (5.14 ± 0.04 μB/f.u.) and 1044.34 kA/m (5.18 ± 0.04 μB/f.u.) at 5 K and 965.00 kA/m (4.79 ± 0.04 μB/f.u.), 967.75 kA/m (4.81 ± 0.04 μB/f.u.), and 1003.52 kA/m (4.97 ± 0.04 μB/f.u.) at 300 K, respectively. The soft ferromagnetic films possess coercivity of ∼ 15.92 kA/m (200 Oe) with a high effective anisotropy constant of ∼ 105 J/m3. Thermo-magnetization measurements yielded high Curie temperatures of 995 K, 1005 K, and 1123 K for films of thicknesses 320 nm, 780 nm, and 1400 nm, respectively. Tailoring the thickness from 320 to 1400 nm could be achieved by merely varying the electrodeposition time without appreciable departure from chemical stoichiometry or compromise in crystalline order or magnetic properties. This establishes a methodology to prepare high quality L21 ordered Co2FeSn films with a few hundred nanometers thickness by electrodeposition. Electrodeposited Co2FeSn films with high magnetic moment, high Keff value, and high Curie temperature are potential candidates for fabricating nanomagnetic devices.

Evaluation of neutronic characteristics of accident tolerant fuel concepts in SMART reactor fuel assemblies using DRAGON

The selection of perfect accident-tolerant nuclear reactor fuels necessitates a neutronic study of the various fuels for Small Modular Reactors (SMRs). This study demonstrates a neutronic assessment of 17 × 17 fuel assemblies of Korean SMART (System-Integrated Modular Advanced ReacTor) integral type SMR using three types of Accident Tolerant Fuel (ATF) by the deterministic code DRAGON. These reactors are less expensive to build, have a shorter construction period, are safer, and have a higher power density. Uranium Nitride (UN), Uranium Di-Silicide (U3Si2), and Composite material (70 % UN- 30 % U3Si2) are considered the most potential ATF candidates. The effective multiplication factor (keff), different fissionable materials, neutron poisons, neutron-induced power, decay power, and fuel burnup are investigated as well as presented on graphs and compared from the conventional fuel UO2. Throughout the entire duration of the operation, all four models maintain a critical state while maintaining consistent dimensions. However, due to enrichment and chemical composition differences, it is found that the UN has the highest keff value. UN displayed lower concentration variation for different isotopes, whereas UO2 showed the highest power analysis study among the models. The different approaches to Accident Tolerant Fuel are not quite significant. However, opting for Uranium Nitride (UN) could offer a slight advantage in terms of effectiveness because of the higher power-up rates and burnup as well as increased cycle length when compared to the other three options.

Validation of OpenMC Code Criticality Value Calculation for GFR Reactor with UN-PuN Fuel

The use of OpenMC code needs to be validated with other code, so that an accurate and valid criticality value is generated. The validation value is affected by the minimum number of particles used. Determination of the minimum number of particles is carried out by varying the number of particles so that they have a convergent value, which is in the range of 100-50000 particles. The results of determining the minimum number of particles are then used to find the minimum number of cycles consisting of active and inactive batches. This study uses the criticality value parameter in the form of an effective multiplication factor (k-eff) as a benchmark for code accuracy. The k-eff values generated by OpenMC and SRAC are then compared and the validation error value is searched. The error value is found by calculating the k-eff (Δk-eff) difference between the two codes. OpenMC code can be said to be validated if it has an error value of less than 1% against the calculation results of the SRAC code. The calculation results of determining the minimum number of particles show k-eff values in the range of 35000 - 50000 minimal changes significantly or not fluctuating. The calculation of k-eff and entropy in cycle 500 shows a convergent value at the active batch value of >30. Based on the data cycle, researchers used 100 active batches and 30 inactive batches to perform validation calculations. The results of the validation calculation using UN-PuN fuel with a plutonium material composition of 10% showed a maximum k-eff error value of 0.819%. The Δk-eff obtained is 0.008.

Effective Multiplication Factor Analysis of Gas-cooled Reactor using OpenMC code with ENDF/B-VII.1, ENDF/B-VIII.0 and JENDL-5.0 Nuclear Data

As the demand for nuclear energy keeps rising, Generation IV reactor designs continuously improve. One such potential reactor design is the Gas-cooled Reactor. Neutronic analysis is the foundation for ongoing design modifications and alternatives for the Helium Gas-cooled Reactor. An important aspect of neutronic analysis is the nuclear data library. This research compares how different nuclear data libraries impact a Helium Gas-cooled Reactor design by comparing the neutron effective multiplication factor (k-eff) value. This research was conducted by the Monte Carlo method using OpenMC code. ENDF/B-VII.1, ENDF/B-VIII.0 and JENDL-5.0 were used as nuclear data libraries to simulate the reactor design for ten years of depletion time with 1-year timesteps. The reactor is designed in two different spectrums: the fast neutron spectrum design and the thermal neutron spectrum design. The differences between the two designs are in the cladding and reflector materials where the thermal neutron spectrum uses graphite moderator. The results for thermal spectrum design show that while each nuclear data library’s average k-eff values vary, they all produce the same results at timestep zero, with k-eff values around 1.02500. The average k-eff value from simulation using the ENDF/B-VII.1 library is 1.01477, the average k-eff value from simulation using the ENDF/B-VIII.0 library is 1.01122, and the average k-eff value from simulation using the JENDL-5.0 library is 1.01047. Each simulation result has an average uncertainty of 0.00040. The difference is due to the different amounts of data in each library. JENDL-5.0 has the highest amount of data inside its nuclear libraries, followed by ENDF/B-VIII.0 and ENDF/B-VII.1. Meanwhile, the average difference between the different libraries is insignificant for the thermal gas-cooled reactor simulation results. ENDF/B-VIII.0 produces an effective k-average of 1.04437, and JENDL-5.0 produces an effective k-average of 1.04273. Mass change over time results also shows that the transmutation of Uranium-238 into Plutonium-239 occurred. The fast spectrum reactor design shows a greater increase in plutonium-239 than the thermal spectrum design. The three libraries used did not show large differences in results in calculating changes in the mass of Uranium-238 over time due to Uranium-238‘s large mass, but they still showed differences in Plutonium-239‘s mass change over time. The average value difference between libraries in the fast neutron spectrum reactor design is 50 kg, and in the thermal spectrum it is 30 kg.

A comparative analysis of gas-cooled fast reactor using heterogeneous core configurations with three and five fuel variations

GFR or Gas-cooled Fast Reactor is one type of fast generation-IV that uses a very high cooling temperature. Thus, it is necessary to have the right reactor core design so that the power distribution of neutrons produced reaches a safe and even limit point. The use of a uniform (homogeneous) reactor core can produce peaking power. This is very avoidable because it will cause a reactor accident. In this study, researchers tried to compare the results of the analysis for two heterogeneous reactor core designs including the configuration of 3 fuel variations and 5 fuel variations using UN-PuN fuel. This study aims to determine the keff value produced by both types of fuel variations during 5 years of burn-up and determine the characteristics of neutron flux, fission rate, and fission product during 15 years of burn-up. This study was started by calculating the homogeneous and heterogeneous core of 3 and 5 fuel variations with neutron transport simulation involving OpenMC. The calculation results show that the heterogeneous core configuration of 5 fuel variations for the keff value is more optimal than 3 fuel variations, because it has the smallest excess reactivity value. The neutron flux and fission rate characteristics for 5 fuel variations are more evenly distributed when compared to 3 fuel variations to maintain neutron lifetime and reactor life in operation. Burn-up residual plutonium material and minor actinide waste for 5 fuel variations have less mass than 3 fuel variations. The results of neutronic analysis of GFR reactors with heterogeneous reactor core designs for 5 fuel variations are better in terms of reactor criticality, neutron power distribution, and waste produced. Finally, optimization of the UN-PuN fuel volume fraction of 60 % provides the optimal keff value

A novel numerical method to evaluate the heat transfer characteristics of complicated CICC structures

In this article, we propose a numerical model for calculating the external heat transfer coefficient between the bundle region and its wall for cables in conduit conductor (CICC). With the assumption of local thermal equilibrium, one macro equation describing the steady heat transfer on the cross section of the CICC can be obtained. A key parameter, the effective transverse thermal conductivity keff, which takes the contribution of both strands and flowing fluid into consideration, is introduced. To calculate the effective transverse thermal conductivity keff, we first obtain the true distribution of different phases (fluid, copper and superconducting material) on a cross section of CICC with the help of the image recognition technique. Based on this, the value of the effective transverse thermal conductivity keff can be calculated numerically. Due to the large difference among the component thermal conductivities (at 4.5 K, typical values of the thermal conductivity of liquid helium, superconducting material and copper are kHe=0.024Wm−1K−1, kNb3Sn=0.04Wm−1K−1 and kCu=708Wm−1K−1; and the maximal ratio of thermal conductivity can be as high as about 30,000), the high-performance finite analytical method (FAM) is recommended to calculate keff. After obtaining the effective transverse thermal conductivity keff of the bundle region, the heat transfer coefficient can be calculated directly by solving a simple Poisson equation under proper boundary conditions. Two case studies are performed, including the JT-60 Super Advanced (JT-60SA) CICC and the dual channel International Thermonuclear Experimental Reactor, Toroidal Field Performance Sample (ITER-TFPS). The calculated values of heat transfer coefficient are consistent with the experimental results, which verifies our proposed model.

Utilizing discarded face masks to fabricate sustainable high-performance panels for enhanced building thermal and acoustic comfort

This study aims to design, fabricate, and optimize fibrous panels made of discarded polypropylene face masks (FMs) for application in sound absorption and thermal insulation. The discarded FMs were shredded and transformed into a fibrous mass using a Garnet machine. The resulting fibers, after being mixed with a bio-based binder, were molded based on an experimental design using response surface methodology in conjunction with the central composite experiment design (RSM-CCD) with five different thicknesses, densities, and blending ratios of spunbond to meltblown fabrics. The average of the sound absorption coefficients for the twelve one‐third‐octave bands from 200 to 2500 Hz (SAA) and effective thermal conductivity (Keff) of the panels were measured as the dependent parameters. These values were calculated within the ranges of 0.28–0.53 and 0.031–0.056 W/(mK), respectively. Additionally, the interactive effects of input variables on SAA and Keff of the manufactured panels were discussed. The quadratic and two-factor interaction models were proposed for optimizing the values of SAA and Keff, respectively. The optimal conditions were determined to achieve the maximum SAA and minimum Keff in a thickness of 26.7 mm, density of 135 kg/m3, and blending percentage of 32.4 %. Under these conditions, the measured values of SAA and Keff were 0.534 and 0.0279 W/mK, respectively. Finally, three fibrous panels were manufactured under optimal conditions, and their acoustic and thermal properties were measured. It was observed that excellent agreement exists between the model results and the experimental data with a 95% confidence level. The bending properties of the optimally designed panels were also investigated and the panels exhibited an impressive flexural strength of 2.1 MPa. Additionally, the fire performance of the panels was examined and all panels showed self-extinguishing properties. These findings suggest that panels made of discarded FMs hold promise as an eco-friendly and sustainable construction material for manufacturing acoustic and thermal insulation panels. Moreover, the frequency-dependent absorption coefficients of the panels were predicted using the Johnson-Champoux-Allard model.

Analysis of variation minor actinide pin configurations Np-237, AM-241, and Cm-244 in UN-PuN fueled pressurized water reactor

Actinide minor is a reactor waste with high toxicity and a long half-life. Minor actinides can be reduced by reusing them as fuel mixtures in reactors. This research uses PWR reactors with the primary fuel UN-PuN or Uranium Plutonium Nitride with a burning time of 5 years. The fuel consists of enriched Uranium, reactor-grade Plutonium from LWR waste, and minor actinides including Neptunium-237, Americium-241, and Curium-244. The purpose of this study was to find a design that is effective in reducing minor actinide waste. There are six designs or cases used in the addition of minor actinides. Each case has six minor actinide pins in each assembly. The addition of minor actinides is arranged in heterogeneous cores. The analysis was carried out by observing the values of k-eff, excess reactivity, and mass of minor actinides obtained from simulations using OpenMC code 0.13.2 and the ENDF/B-VIII library. The homogeneous core obtained an excess reactivity of 9.7 % with a percentage of plutonium of 8 %. The results of the homogeneous core are used as a reference for preparing a heterogeneous core. The heterogeneous core obtained an excess reactivity of 9.9 % with a percentage of plutonium F1: 5.5 %, F2: 8 %, and F3: 10.5 %. Np-237 can be reduced by 53 kg, and Am-241 can be reduced by 61 kg with minor actinide pins in case 1. Cm-244 can be reduced by 363 kilograms with minor actinide pins in case 6. Excess reactivity in the addition of Np-237 and Am-241 decreased to 5.3 %, while the accumulation of Cm-244 increased to 12.1 %.

Epoxy‐hemp and epoxy‐flax composites filled with glass dust for enhanced thermal insulation: An analytical and experimental study

AbstractThis article reports on an analytical and experimental investigation on the effects of fiber and filler reinforcement on the effective thermal conductivity (keff) of epoxy‐based composites. This research aims at enhancing the thermal insulation capability of fiber reinforced composites by adding an industrial waste, WGD (waste glass dust) as particulate filler. For this, an analytical model is proposed to estimate the keff values of particle filled polymer composites with and without fiber reinforcement. The results computed from the proposed correlations are then validated with some previously reported investigations. For an experimental validation, hemp and flax fiber reinforced epoxy based hybrid composites are fabricated with WGD as the fillers in varying weight fractions of 0 to 0.2. Effects of filler content on keff of fiber‐polymer systems are studied. Their keff are measured by a Unitherm Model 2022 instrument. It is found that the predicted and measured values are in good agreement. It is observed that the keff of the epoxy composites reduced from 0.359 to 0.297 W/m K with increase in the WGD weight fraction from 0 to 0.2. The keff of hybrid composites are also reduced by 27% and 15.53%, respectively, with 20 wt% addition of WGD.Highlights Hybridization of natural fiber based polymer composites. Utilization of waste glass dust as functional filler. Model development for heat conduction across hybrid composite system. Insulation capability enhancement of epoxy using fiber/filler reinforcements. Low cost thermal insulation materials using green materials and industry waste.