المؤلفون: Insley, Benjamin¹ (AUTHOR) BAInsley@mdanderson.org, Bartkoski, Dirk² (AUTHOR), Balter, Peter¹ (AUTHOR), Prajapati, Surendra¹ (AUTHOR), Tailor, Ramesh¹ (AUTHOR), Jaffray, David³ (AUTHOR), Salehpour, Mohammad¹ (AUTHOR)

المصدر: Physics in Medicine & Biology. 5/21/2024, Vol. 69 Issue 10, p1-15. 15p.

مصطلحات موضوعية: *COLLIMATORS, *NEUTRON transport theory, *PHOTON flux, *SIMULATED annealing, *IRRADIATION, *X-ray tubes, *PHASE space, *LINEAR accelerators

مستخلص: Objective. A novel x-ray field produced by an ultrathin conical target is described in the literature. However, the optimal design for an associated collimator remains ambiguous. Current optimization methods using Monte Carlo calculations restrict the efficiency and robustness of the design process. A more generic optimization method that reduces parameter constraints while minimizing computational load is necessary. A numerical method for optimizing the longitudinal collimator hole geometry for a cylindrically-symmetrical x-ray tube is demonstrated and compared to Monte Carlo calculations. Approach. The x-ray phase space was modelled as a four-dimensional histogram differential in photon initial position, final position, and photon energy. The collimator was modeled as a stack of thin washers with varying inner radii. Simulated annealing was employed to optimize this set of inner radii according to various objective functions calculated on the photon flux at a specified plane. Main results. The analytical transport model used for optimization was validated against Monte Carlo calculations using Geant4 via its wrapper, TOPAS. Optimized collimators and the resulting photon flux profiles are presented for three focal spot sizes and five positions of the source. Optimizations were performed with multiple objective functions based on various weightings of precision, intensity, and field flatness metrics. Finally, a select set of these optimized collimators, plus a parallel-hole collimator for comparison, were modeled in TOPAS. The evolution of the radiation field profiles are presented for various positions of the source for each collimator. Significance. This novel optimization strategy proved consistent and robust across the range of x-ray tube settings regardless of the optimization starting point. Common collimator geometries were re-derived using this algorithm while simultaneously optimizing geometry-specific parameters. The advantages of this strategy over iterative Monte Carlo-based techniques, including computational efficiency, radiation source-specificity, and solution flexibility, make it a desirable optimization method for complex irradiation geometries. [ABSTRACT FROM AUTHOR]

المؤلفون: Kostal, Michal¹ (AUTHOR) michal.kostal@cvrez.cz, Matěj, Zdeněk² (AUTHOR), Schulc, Martin¹ (AUTHOR), Losa, Evžen¹ (AUTHOR), Šimon, Jan¹ (AUTHOR), Novák, Evžen¹ (AUTHOR), Cvachovec, František² (AUTHOR), Přenosil, Vaclav² (AUTHOR), Mravec, Filip² (AUTHOR), Czakoj, Tomáš¹ (AUTHOR), Rypar, Vojtěch¹ (AUTHOR), Trkov, Andrej³ (AUTHOR), Capote, Roberto⁴ (AUTHOR)

المصدر: Nuclear Science & Engineering. Feb2024, Vol. 198 Issue 2, p399-410. 12p.

مصطلحات موضوعية: *STAINLESS steel, *NEUTRON transport theory, *NEUTRONS, *LEAKAGE, *RESEARCH institutes, *NEUTRON sources

مستخلص: Integral experiments covering neutron leakage from geometrically simple assemblies with a 252Cf source inside are very valuable tools usable in the validation of transport cross-section data since geometric uncertainties play a much smaller role in simple geometric assemblies than in complex assemblies as for example reactor pressure vessel geometry. Since 252Cf spontaneous fission is a standard neutron source, the uncertainties connected with the source neutron spectrum can be even neglected. The paper refers to validation efforts of neutron leakage from an ~50 × 50 × 50-cm stainless steel block in the Research Center Rez. Both the neutron leakage flux at a distance of 1 m from the center of the cubical assembly using stilbene spectrometry and activation rates at different positions of the assembly were evaluated. In addition to experiments, main sources of uncertainty were identified and evaluated. The results of the stilbene measurements are consistent with the activation measurement results. [ABSTRACT FROM AUTHOR]

المؤلفون: Favorite, Jeffrey A.¹ (AUTHOR) fave@lanl.gov

المصدر: Nuclear Science & Engineering. Feb2024, Vol. 198 Issue 2, p287-299. 13p.

مصطلحات موضوعية: *NUCLIDES, *NEUTRON transport theory, *NEUTRON sources, *DENSITY

مستخلص: Application of perturbation capabilities for density sensitivities in Monte Carlo radiation transport codes has been limited because changing source nuclide densities or source material densities changes the intrinsic source, and in most Monte Carlo codes, the user-input source is independent of the user-input materials. The perturbation capability then has no way of accounting for changes in the intrinsic source. This paper derives the sensitivity of a response with respect to a source nuclide density in terms of a portion due to the transport operator and a portion due to the source rate density. The Monte Carlo perturbation method computes the portion due to the transport operator, and the portion due to the source rate density is computed in postprocessing using parameters from the precomputed intrinsic source calculation. This paper derives first- and second-order sensitivities. The equations require the response to be separated by contribution from each of the sources modeled. A test problem containing several (α,n) and spontaneous fission neutron sources verifies the method. [ABSTRACT FROM AUTHOR]

المؤلفون: Yin, Shao-Chong, Yu, Jing-Xin, Hou, Zhi-Yu, Liu, Xiu-Ying, Li, Xiao-Dong

المصدر: Journal of Applied Physics; 11/14/2023, Vol. 134 Issue 18, p1-9, 9p

مصطلحات موضوعية: NICKEL electrodes, SPIN polarization, DENSITY functionals, MANGANESE, DENSITY functional theory, NEUTRON transport theory, GREEN'S functions, HEUSLER alloys

مستخلص: In this paper, we investigated the spin transport properties of binuclear manganese phthalocyanine (M n 2 P c 2) spintronic devices sandwiched between two nickel electrodes using the non-equilibrium Green's function method in combination with density functional theory. Based on the calculation results, the M n 2 P c 2 device exhibited excellent spin-filtering capabilities, demonstrating an exceptionally high spin filter efficiency (SFE). Irrespective of the parallel or antiparallel orientation of magnetization in the electrodes, we observed that when both manganese atoms were in a spin-up state, the SFE of spin-resolved currents under finite bias and the thermoelectric currents induced by temperature gradients at fixed temperatures were both close to 100%. The large spin Seebeck polarization of the M n 2 P c 2 device was also obtained at low reference temperatures. This study explores the potential for developing multifunctional spintronic single-molecule devices using Ni − M n 2 P c 2. [ABSTRACT FROM AUTHOR]

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المؤلفون: Schmeissner, J.^1,2 (AUTHOR) yokhan.schmeissner@itep.ru, Tyulyusov, A. N.^1,2 (AUTHOR)

المصدر: Physics of Atomic Nuclei. Dec2023, Vol. 86 Issue 10, p2256-2259. 4p.

مصطلحات موضوعية: *NEUTRON spectrometers, *NEUTRON transport theory, *ABSORPTION cross sections, *CRYSTALS, *GEOMETRY

مستخلص: Calculated rocking curves are presented for the spectrometric scheme of a double-crystal diffractometer made of InSb crystals in the Laue–Laue geometry for neutron wavelength ranges corresponding to the weak potential and strong resonant absorptions. The appearance and shape of the curves calculated using the extended expression with allowance for the absorption cross section and the curves obtained using the Compton–Alisson expression are compared. [ABSTRACT FROM AUTHOR]

المؤلفون: Dawn, William C.¹ (AUTHOR) william.dawn@studsvik.com, Palmtag, Scott¹ (AUTHOR)

المصدر: Nuclear Science & Engineering. Dec2023, Vol. 197 Issue 12, p3138-3159. 22p.

مصطلحات موضوعية: *TRANSPORT equation, *NEUTRON transport theory, *NEUTRONS, *NEUTRON flux, *GRAPHICS processing units, *FINITE element method

مستخلص: The Microreactor Exascale eZ CALculation (MEZCAL) tool has been developed to accurately and efficiently solve the neutron transport equation in general, unstructured meshes to support the design and modeling of microreactors. MEZCAL solves the self-adjoint angular flux form of the neutron transport equation using the finite element method. As the neutron transport equation is computationally expensive to solve, MEZCAL is designed to efficiently use exascale computing architectures, with an emphasis on graphics processing unit computing. To leverage existing tools, MEZCAL is built using the MFEM library and uses solvers from HYPRE, PETSc, and SLEPc. Verification of the neutron transport solver in MEZCAL is demonstrated with the solution to a one-dimensional cylindrical problem that has a semi-analytic solution. After verification, a realistic microreactor based on the MARVEL microreactor design is modeled using MEZCAL. Spatial and angular refinement results are presented for a two-dimensional model of the MARVEL microreactor, and the eigenvalue is converged to approximately 60 pcm. This convergence required a very fine mesh and more than 3.76 Billion Degrees Of Freedom (BDOF). Preliminary results are also presented for a three-dimensional model of the MARVEL microreactor. Finally, a weak scaling study is performed to investigate how the methods in MEZCAL will scale for larger problems with the next generation of exascale computing architectures. [ABSTRACT FROM AUTHOR]

المؤلفون: Tuya, Delgersaikhan¹ (AUTHOR) nagaya.yasunobu@jaea.go.jp, Nagaya, Yasunobu¹ (AUTHOR)

المصدر: Journal of Nuclear Engineering (JNE). Dec2023, Vol. 4 Issue 4, p691-710. 20p.

مصطلحات موضوعية: *FEEDFORWARD neural networks, *NEUTRON transport theory, *PROBABILITY density function, *MONTE Carlo method, *CONTINUOUS distributions, *NEUTRON flux, *PROBABILITY theory

مستخلص: The Monte Carlo neutron transport method is used to accurately estimate various quantities, such as k-eigenvalue and integral neutron flux. However, in the case of estimating a distribution of a desired quantity, the Monte Carlo method does not typically provide continuous distribution. Recently, the functional expansion tally (FET) and kernel density estimation (KDE) methods have been developed to provide a continuous distribution of a Monte Carlo tally. In this paper, we propose a method to estimate a continuous distribution of a quantity in all phase-space variables using a fully connected feedforward artificial neural network (ANN) model with Monte Carlo-based training data. As a proof of concept, a continuous distribution of iterated fission probability (IFP) was estimated by ANN models in two distinct fissile systems. The ANN models were trained on the training data created using the Monte Carlo IFP method. The estimated IFP distributions by the ANN models were compared with the Monte Carlo-based data that include the training data. Additionally, the IFP distributions by the ANN models were also compared with the adjoint angular neutron flux distributions obtained with the deterministic neutron transport code PARTISN. The comparisons showed varying degrees of agreement or discrepancy; however, it was observed that the ANN models learned the general trend of the IFP distributions from the Monte Carlo-based training data. [ABSTRACT FROM AUTHOR]

المؤلفون: Zhang, Dingkang¹ (AUTHOR), Rahnema, Farzad¹ (AUTHOR) farzad@gatech.edu

المصدر: Nuclear Science & Engineering. Sep2023, Vol. 197 Issue 9, p2498-2508. 11p.

مصطلحات موضوعية: *NEUTRON transport theory, *COMETS, *NEUTRON scattering, *NUCLEAR reactor cores, *BENCHMARK problems (Computer science), *NEUTRON sources

مستخلص: The COarse MEsh Transport (COMET) method, a hybrid continuous energy stochastic and deterministic transport method/tool based on the incident flux response expansion theory, is capable of providing highly accurate and efficient continuous energy whole-core neutron solutions to various heterogeneous reactor cores. In this work, a novel low-order (zeroth-order) acceleration technique is developed to significantly improve COMET's computational efficiency for core calculations. This new method is based on consistent coupled low-order and high-order calculations to obtain the COMET core solution. In the low-order calculations, COMET is used to converge the total partial current escaping from each coarse mesh and the core eigenvalue. The resulting fixed-source problem in which the off-diagonal terms (equivalent to the scattering and fission neutron sources) are constructed by the zeroth-order solution are efficiently solved by the high-order COMET calculations. The resulting high-order angular flux on each coarse mesh bounding surface is then used to update (collapse) the low-order response coefficients. The coupled low-order and high-order calculations are repeated until both the eigenvalue and the low-order response coefficients are converged. The new acceleration method is implemented into COMET and tested in a set of stylized Advanced High Temperature Reactor (AHTR) benchmark problems. It is found that the core eigenvalues and the local fission density distributions predicted by COMET with the low-order acceleration agree very well with those computed by the original COMET. The eigenvalue discrepancy varies from 0 to 1 pcm, and the average relative differences in the stripewise and assembly-average fission density distributions are in the range of 0.021% to 0.032% and 0.004% to 0.01%, respectively. The comparisons have shown that the new low-order acceleration method can maintain COMET's accuracy while improving its computational efficiency for core calculations by 12 to 16 times. [ABSTRACT FROM AUTHOR]

المؤلفون: Grigorova, Vili¹ (AUTHOR) vili.grigorova@mq.edu.au, Shcheka, Svyatoslav² (AUTHOR), Salehi, Fatemeh¹ (AUTHOR), Kamenev, Konstantin V.³ (AUTHOR), Clark, Simon M.¹ (AUTHOR)

المصدر: High Pressure Research. Sep2023, Vol. 43 Issue 3, p215-230. 16p.

مصطلحات موضوعية: *FINITE element method, *NEUTRON transport theory, *THERMAL analysis, *NEUTRONS, *TOMOGRAPHY, *TEMPERATURE distribution

مصطلحات جغرافية: PARIS (France)

مستخلص: A 3D Finite Element Analysis model was developed to describe the temperature distribution inside a novel neutron transparent high-pressure sample assembly. The validity of the model was established by conducting experimental validation. The Finite Element Analysis model was utilised to evaluate the sample assembly's temperature gradients and optimise its geometry and components. The results indicate that the discrepancy between the temperature recorded in the laboratory using thermocouples and the temperature calculated by the Finite Element Analysis model was only 4% under the assumption that the thermocouple was positioned in the middle of the sample assembly. The models further demonstrate that the Finite Element Analysis approach is a valuable tool for optimising the sample assembly by considering the impact of different materials and variations in the shape of its components. [ABSTRACT FROM AUTHOR]

المؤلفون: Assogba, Kenneth^1,2 (AUTHOR) kenneth.assogba@cea.fr, Bourhrara, Lahbib¹ (AUTHOR), Zmijarevic, Igor¹ (AUTHOR), Allaire, Grégoire² (AUTHOR), Galia, Antonio¹ (AUTHOR)

المصدر: Nuclear Science & Engineering. Aug2023, Vol. 197 Issue 8, p1584-1599. 16p.

مصطلحات موضوعية: *NEUTRON transport theory, *SPHERICAL harmonics, *TRANSPORT equation, *FINITE element method, *NEUTRONS, *SPARSE matrices

مستخلص: The spherical harmonics or PN method is intended to approximate the neutron angular flux by a linear combination of spherical harmonics of degree at most N. In this work, the PN method is combined with the discontinuous Galerkin (DG) finite elements method and yield to a full discretization of the multigroup neutron transport equation. The employed method is able to handle all geometries describing the fuel elements without any simplification nor homogenization. Moreover, the use of the matrix assembly-free method avoids building large sparse matrices, which enables producing high-order solutions in a small computational time and less storage usage. The resulting transport solver, called NYMO, has a wide range of applications; it can be used for a core calculation as well as for a precise 281-group lattice calculation accounting for anisotropic scattering. To assess the accuracy of this numerical scheme, it is applied to a three-dimensional (3-D) reactor core and fuel assembly calculations. These calculations point out that the proposed PN -DG method is capable of producing precise solutions, while the developed solver is able to handle complex 3-D core and assembly geometries. [ABSTRACT FROM AUTHOR]