Results 1 - 10 of 377
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[en] Highlights: •A modified α − k power iteration method is presented for computing the time-eigenvalue. •It is not required to provide the initial values of α for the modified method. •Computational experiences validate the validity and efficiency of the new method. -- Abstract: A modified α − k power iteration method is presented for the prediction of time-eigenvalue(α) of the neutron transport equation. By developing a direct relationship between K-eigenvalue and α-eigenvalue, a new formula is introduced to estimate the value of α. Compared with the conventional method, it is not required to provide the initial values of α for the modified method. Since it is always difficult to guess the suitable initial values, the modified method is more convenient for solving time-eigenvalue problems. Computational experiences show that the accuracy of the modified method is the same as the conventional method.
[en] Nuclear reactor physics deals with the solution of the neutron transport equation, for which several numerical methods and resulting calculation codes exist. Using parallel programming to multiply the resources available to these codes, it is possible to increase their computational power, thus improving their modelling capabilities. This work focuses on the optimization of neutron transport algorithms by means of parallel programming techniques. For cell calculations, the collision probabilities and heterogeneous response (HRM) methods are considered, including the associated multigroup scheme, as implemented in INVAP’s cell calculation code, CONDOR. The diffusion method in a finite difference formulation, used in CITVAP for core calculations, is optimized as well. The parallel methods obtained are implemented in OpenMP, an application for parallel programming in shared memory computers. (author)
[es]La física de reactores está basada en la resolución de la ecuación de transporte de neutrones, para la cual existen diversos esquemas numéricos y códigos de cálculo que los implementan. La programación en paralelo permite aumentar la velocidad de estos códigos mediante la utilización de múltiples unidades de procesamiento, mejorando la capacidad de modelado. En este trabajo se estudió la paralelización de distintos algoritmos asociados a la ecuación de transporte. Para la etapa de cálculos de celda fueron considerados los métodos de probabilidades de colisión y de respuesta heterogénea (HRM), incluyendo el esquema multigrupo asociado, y su implementación en el código CONDOR, desarrollado en INVAP. El método de difusión en diferencias finitas implementado en el código CITVAP fue analizado para los cálculos de núcleo. La programación de los algoritmos desarrollados se realizó en OpenMP, una herramienta de procesamiento en paralelo en el modelo de memoria compartida. (author)
[en] We investigate transport theory for anisotropic transport of neutrons in finite medium or injected externally. When anisotropic transport is treated by the usual transport equation, on which reversibility of collisions is shown imposed, successive collisions always induce 'self-collision' or sham collision; the fact is unavoidable as long as statistical ensemble is constructed from the reductionistic mechanical-systems. Then, irreductionistic elements, or spatial cells containing assembly of free neutrons (and implicit medium nuclei) uniformly are introduced, from which alternative Liouville equation is constructed. Successive collisions are expressed by fusing three cells; for reviving mechanical law in the collisions the law of action and reaction is applied to between first fused-cell and third cell. Extended transport equation can thus describe the process of chaotically mixing anisotropic momentum, i.e., the well-known deep penetration. (author)
[en] In order to simulate the complex structure fuel assembly, Nuclear Power Institute of China(NPIC) has developed an advanced neutron transport lattice code, named KYLIN-2. The subgroup method is adopted in KYLIN-2 to treat resonance problems. And method of characteristics(MOC) is used to solve neutron transport equation and generalized coarse mesh finite difference(GCMFD) method is used to accelerate the calculation. In order to solve the depletion equation, CRAM method and PPC method are used. The graphical input and display interface are developed to make sure the KYLIN-2 code can be used easily by engineers. The numerical results show that the KYLIN-2 can calculate the PWR assembly accurately. (authors)
[en] This paper is concerned with the validation of the 3D deterministic neutral-particle transport theory code EVENT for shielding applications. The code is based on the finite element-spherical harmonics (FE-PN) method which has been extensively developed over the last decade. A general multi-group, anisotropic scattering formalism enables the code to address realistic steady state and time dependent, multi-dimensional coupled neutron/gamma radiation transport problems involving high scattering and deep penetration alike. The powerful geometrical flexibility and competitive computational effort makes the code an attractive tool for shielding applications. In recognition of this, EVENT is currently in the process of being adopted by the UK nuclear industry. The theory behind EVENT is described and its numerical implementation is outlined. Numerical results obtained by the code are compared with predictions of the Monte Carlo code MCBEND and also with the results from benchmark shielding experiments. In particular, results are presented for the ASPIS experimental configuration for both neutron and gamma ray calculations using the BUGLE 96 nuclear data library. (author)
[en] JMCT is a general purpose Monte Carlo neutron-photon-electron or coupled neutron/photon/electron transport code with a continuous energy and multigroup. The code has almost all functions of a general Monte Carlo code which include the various variance reduction techniques, the multi-level parallel computation of MPI and OpenMP, the domain decomposition and on-fly Doppler broadening, etc. Especially, JMCT supports the depletion calculation with TTA and CRAM methods. The input uses the CAD modelling and the calculated results use the visual output. The geometry zones, materials, tallies, depletion zones, memories and the period of random number are enough big for suiting various problems. This paper describes the application of the JMCT Monte Carlo code to the simulation of BEAVRS and SG-III shielding model. For BEAVRS model, the JMCT results of HZP status are almost the same with MC21, OpenMC and experiment. Also, we performed the coupled calculation of neutron transport and depletion in full power. The results of ten depletion steps are obtained, where the depletion regions exceed 1.5 million and 120 thousand processors to be used. Due to no coupled with thermal hydraulics, the result is only for reference. Finally, we performed the detail modelling for Chinese SG-III laser facility, where the anomalistic geometry bodies exceed 10 thousands. The flux distribution of the radiation shielding is obtain based on the mesh tally in case of Deuterium-Tritium fusion reaction. The high fidelity of JMCT has been shown. (authors)
[en] Highlights: • A GPU-based parallel MOC algorithm is implemented, which includes several solving kernels. • A performance analysis model is applied to analyze the performance of the code and identify the limitation. • The corresponding optimizations according to the analysis are applied and the significant speedup ratio is obtained. - Abstract: The method of characteristics (MOC) is one of the most common methods for solving the neutron transport equation in practical application. Researches have been focused on the acceleration techniques and the parallel algorithm for improving the efficiency of MOC. The Graphics Processing Unit (GPU) provides an alternative method of parallelizing the MOC neutron transport sweep. In this work, a GPU-paralleled 2D MOC code is implemented, which employs the diamond difference (DD) scheme and the step characteristics (SC) scheme. Different parallel schemes which are ray-level, energy-group-level, and polar-angle-level, are analyzed to choose the proper parallel scheme. The C5G7 2D benchmark is calculated to verify the accuracy and efficiency of the code in different schemes with single precision and double precision. The bottlenecks of the GPU code are identified and the code is classified into three categories, which are compute-bound, memory-bound, and latency-bound, according to the performance analysis model introduced in this paper. In addition, corresponding optimization strategies are applied to improve the performance according to the analysis. Moreover, the speed, power efficiency, and hardware cost are compared for CPU and GPU based on a fictitious quarter core PWR problem. Numerical results demonstrate that the energy group-level parallelization can obtain the optimal performance on GPU. Optimization strategies are effective to improve the efficiency of the calculation on GPU, which indicates that the performance analysis model is useful and effective to locate the limitation of the code. Moreover, the GPU-version code is about 30 times faster than the CPU-version code with double precision and about 100 times faster with single precision, while the desired accuracy is kept. And the GPU delivers superior performance in both speed, energy efficiency, and hardware cost.
[en] Highlights: • Synchronized implementation of the method of characteristic (MOC) for neutron transport in forward and adjoint. • Verification and the validation of the numerical scheme without necessity for experimental benchmarks. • Fast and accurate numerical computation which could be extended to full core calculations. - Abstract: Neutron transport adjoint calculations are useful in many reactor physics applications. Among various applications, the adjoint flux can be used in perturbation theory to prevent large calculations and computational costs when estimating a small reactivity insertion into the system. Furthermore, they can serve as validation tests for numerical schemes, since both direct and adjoint calculations for a given system should lead to the same eigenvalue, although using two different physicomathematical formulations of the transport model. The synchronized implementation of the method of characteristic (MOC) for neutron transport in forward and adjoint approaches is accomplished in this work. The result is validated using the C5G7 benchmark with comparisons of the multiplication factor and pin power values. Differences between forward and adjoint multiplication factors in the results are achieved in the order of 1.0E-6. Meanwhile, the difference between the multiplication factor and the C5G7 benchmark is in order of 1.0E-5, using an S16 level symmetric angular discretization and a track spacing of 0.01 cm.