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[en] Highlights: • Thorough analysis on all the cases in C5G7-TD benchmark. • Parametric study on the time step size. • Comprehensive comparison with other codes’ results. - Abstract: Verification and Validation (V&V) serves an important role in assessing the numerical methods implemented in a neutron transport code. However, very few benchmarks are available for verifying and validating the transient capability of neutron transport codes. In this paper, the transient methodology of the Transient Multilevel (TML) Method used in the MPACT code was verified using Phase I of the recently developed OECD/NEA C5G7-TD benchmark, which was specifically designed for validating the transport space-time simulations. The results in MPACT agree well with results of other codes. While this is not code validation, the results contributes to the MPACT verification base and provide an additional solution for the C5G7-TD benchmark which can increase the importance of the OECD benchmark to assess the performance of other transient neutron transport codes.
[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.
[en] This study is concerned with the investigation of the half-space albedo problem for “İnönü-linear-quadratic anisotropic scattering” by the usage of Modified F N method. The method is based on CASE's method. Therefore, CASE's eigenfunctions and its orthogonality properties are derived for anisotropic scattering of interest. Albedo values are calculated for various linear, quadratic and İnönü anisotropic scattering coefficients and tabulated in Tables.
[en] Generally, solving a neutron transport equation on a domain with curved boundaries by using mesh-based methods requires to approximate the domain geometry. This increases the computation time in some specified cases that a very fine regular mesh must be used to approximate geometry. Several unstructured mesh methods have been developed, but all of these methods suffer from limitations related to the shape of the used unstructured mesh. In this study, a new 2D unstructured discrete ordinates method (Sn) is developed to solve discrete ordinates equation on a generally unstructured mesh. By using radial basis function (RBF), it provides a flexible technique for discretizing the spatial variable of the Sn equation without any requirement related to the shape of unstructured mesh as well as the number of spatial dimensions. The work in this study includes a two-step approach. First, the two types of meshfree approximation methods, moving least squares (MLS) and radial basis function (RBF), are investigated for implementation capability and accuracy for solving a diffusion equation in a weak-strong form. Attention is paid to the approximation quality on the boundaries and the stability of the discrete equation system which are affected by support domain of shape function of the MLS and RBF. This step lays the groundwork for developing an unstructured discrete ordinates method based on the RBF approximation. In the developed method, the discretization of the spatial variable in Sn equation is carried out on the polygons which overlap each other. The flux within each polygon is approximated only by nodes which belong to the polygon. It is intended to prevent oscillation in the flux due to Gibb phenomenon which appears when a support radius of the RBF is large. By doing so, it allows the balance the accuracy of discretization, but still retaining the flexibility of unstructured mesh in the shape. Several benchmarks are implemented and compared to the standard Sn method to verify the accuracy and stability of the developed method. The yielded results are in good agreement with the standard Sn results. The discrete equation system based on this approach also achieves the stability in the condition of the average density of solution nodes is not quasi-uniform. Therefore, it is expected that the unstructured discrete ordinates method developed in this study is possible to be implemented in the neutron transport calculation
[en] The radionuclide inventory plays a central role in the safety of nuclear installations both during operation and their decommissioning. In nuclear fusion reactors using Pb–Li tritium breeding blankets, the undesired production of radiotoxic 210Po is still an unresolved safety issue. In this work, neutron transport calculations and inventory calculations are combined to predict the 210Po inventory in a DEMO fusion reactor using either a helium cooled lithium lead or a water cooled lithium lead breeding blanket. In order to guarantee that the environmental 210Po release associated with an ex-vessel leak-of-PbLi accident remains below the no-evacuation limit, the 210Po concentration in the Pb–Li should be kept below 1500 appt. It was found that no Pb–Li purification is required to keep the 210Po concentration in DEMO below this limit. However, in case the Pb–Li makes direct contact with water, more volatile Po-containing (oxy-)hydroxides could form. If these species increase the 210Po release rate by more than a factor of two, safety measures will be required. Therefore, 210Po generation in DEMO does not pose a hazard in case of a regular ex-vessel leak-of-PbLi accident, unless possibly in case the Pb–Li makes contact with water. (paper)
[en] Validation and evaluation of the recently released ADVANTG code, which combines a well-known Monte Carlo (MC) transport code MCNP with a deterministic neutron transport code Denovo, is presented in the paper. The aim of ADVANTG is to automate the process of generating the variance reduction parameters for the fixed source MCNP calculations, which consequently accelerate the simulations in terms of the required CPU time. Reliability and consistent performance of the ADVANTG code were tested on a computationally demanding benchmark, i.e. the skyshine benchmark experiment from the ICSBEP handbook where neutron and photon scattering in the air above an open operating reactor are simulated. The speed-up factors or the increases in relative efficiency of up to 30,000 and 1400 compared to analog MCNP simulations were achieved using the ADVANTG-generated variance reduction parameters for neutrons and photons respectively. As the mean values obtained by the ADVANTG-accelerated simulation sit within the statistical uncertainties of the analog simulation, it was shown that no additional bias is introduced by the ADVANTG code. This article shows that ADVANTG code is a useful and effective tool for use in solving neutral particle transport problems in large geometries featuring skyshine.
[en] The nuclear simulation chain of GRS provides powerful tools in the field of nuclear in-core and ex-core safety analysis. In terms of improved robustness under severe accident conditions, recently diverse innovative fuel materials are under investigation within the concept called 'Accident Tolerant Fuels' (ATF). Here different materials amend or replace the well-known fuel components, introducing new isotopes to the system such as iron, chromium, aluminum, silicon, carbon, or others. Besides the thermo-mechanical properties primarily under scope, some of these materials feature neutronic properties which differ from the standard UO2/Zr system and thus impact on various safety-related nuclear in-core and ex-core characteristics. In addition to new materials, also different arrangements of materials occur, e.g. multilayered cladding materials or heterogeneous fuel forms as TRISO-like particles in an inert matrix forming water-moderated fuel rods. These arrangements pose new challenges to accurate cross-section processing in terms of resonance self-shielding, resonance treatment, and collapsing to few-group cross-sections. As for the overall neutronic calculation chain, this directly affects the problem-dependent reactor core simulations as well as waste management applications. The robust and reliable applicability of the overall GRS nuclear simulation chain to ATF needs to be demonstrated. First results of this ongoing work are shown here, featuring pin-cell and assembly-wise criticality, cross-section processing, and burn-up calculations using different sequences from the SCALE code system with ENDF/B-VII.1 based cross-section libraries. Comparative analyses using the SERPENT code with ENDF/B-VII.0 based continuous energy cross-sections for criticality calculations, few-group cross-section generation for reactor physics applications, and inventory determination are performed. First code-to-code and experimental benchmark exercises are discussed. Reasonable or good agreement between the different calculation systems as well as to the benchmark experiments have been found so far. Hence, up to now no insurmountable obstacles have been observed to accurately account for the peculiarities of some ATF as compared to the standard system. However, for a final evaluation, further investigations need to be performed and assessed. (authors)
[en] Highlights: • This is the first Method of characteristics code for pebble bed reactor. • Special geometry modeling is developed for pebble bed reactor. • Special algorithm for ray tracing is optimized pebble bed reactor. - Abstract: The method of characteristics which is an effective method for neutron transport calculation with high flexibility in complex geometry has been applied to high temperature gas-cooled pebble-bed reactor geometry. The MOCP (the Method of Characteristics for Pebble-bed high temperature gas-cooled reactor) code was developed in Institute of Nuclear and New Energy Technology, Tsinghua University to achieve this goal. The first stone standing ahead is the geometry modeling for pebble-bed district. In this work, a solid modeling method called constructive solid geometry (CSG) is introduced to model the pebble-bed district. A universal ray tracing technique is developed to arrange tracks in 3-D cylinder geometry. Furthermore, tree search method is implemented into MOCP to improve the ray tracing efficiency. Automatic meshing is implemented to improve the geometry modeling efficiency. A redundant tree was constructed to mesh the interspace among pebbles. The geometry modeling capability of MOCP is verified by fast converging single group benchmarks.