Neutron Transport Calculations Research Articles

Observation of neutron emission anisotropy by neutron activation measurement in beam-injected LHD deuterium plasmas

Neutron emission anisotropy caused by neutral beam injection was observed by the neutron activation measurement in deuterium plasmas confined in the Large Helical Device. The shot-integrated fast neutron flux was measured by the irradiation of activation foils at two irradiation ends located at different ports. The ratio of the flux at the outside horizontal port to that at the lower port was used as an index of neutron emission anisotropy. The dependence of the ratio on the direction of the neutral beam injection was confirmed by comparing cases of tangential, perpendicular and both tangential and perpendicular neutral beam injections. Neutron emission anisotropy was numerically evaluated assuming these three cases of neutral beam injection. The analysis of neutron emission anisotropy comprises the evaluation of the velocity distribution function of energetic deuterons following guiding-centre orbits, calculation of the double-differential emission spectrum of neutrons produced by the reaction and neutron transport calculation. The obtained numerical results are qualitatively consistent in the dependence of neutron emission anisotropy on the neutral-beam-injection direction with the observed experimental data.

A Linear Prolongating Coarse Mesh Finite Difference Acceleration of Discrete Ordinate Neutron Transport Calculation Based on Discontinuous Galerkin Finite Element Method

The recently developed linear prolongation Coarse Mesh Finite Difference (lpCMFD) acceleration scheme, which employs a linear additive approach to update the scalar flux, has been shown to be more stable and effective than the conventional scaling-based Coarse Mesh Finite Difference (CMFD) method for accelerating the discrete ordinates (SN) neutron transport calculation using spatial finite difference discretization. In this paper, we study and extend the application of lpCMFD to accelerate the SN neutron transport calculation with spatial discretization using the Discontinuous Galerkin Finite Element Method (DGFEM), which generally involves linear- or higher-order space expansion functions. A function space mapping operator is proposed in this paper to project the lpCMFD linear-order correction flux to an arbitrary-order DGFEM basis function, which is implemented and tested on a one-dimensional (1-D) in-house–developed DGFEM-based SN code. The consistency between the lpCMFD accelerated results and the pure SN results is naturally guaranteed by employing upwind current information from DGFEM-based SN transport calculation to evaluate the drift coefficient. It was found from our numerical testing with the CMFD and the lpCMFD acceleration schemes on single-group fixed-source and k-eigenvalue problems that both acceleration schemes can reproduce the unaccelerated scalar flux and keff, respectively. Further numerical testing on a more realistic case is performed on a 1-D slice multi-energy-group problem based on the three-dimensional C5G7 mixed oxide (MOX) benchmark. It was found that by using the function space projector proposed in this paper, lpCMFD was stable and effective to accelerate the DGFEM-based SN neutron transport calculation for all coarse mesh sizes tested while CMFD diverged for large optical thickness.

Three-dimensional method of characteristics/diamond-difference transport analysis method in STREAM for whole-core neutron transport calculation

A new three-dimensional (3D) transport analysis method is developed and implemented in the light water reactor (LWR) whole-core analysis code STREAM. The method is named as 3D method of characteristics/Diamond-difference (MOC/DD). In this method, 3D variables such as flux and source are expressed as a combination of the two-dimensional (2D) radial component and the one-dimensional (1D) axial component. The 2D method of characteristics (MOC), which is widely used in 2D neutron transport analysis in reactor physics, is used to solve complicated radial geometries composed of fuel pin, burnable absorber, structural material, etc. The axial neutron source, which is discretized by the diamond-difference (DD) method, is merged into the 3D MOC source term. This paper presents a detailed derivation of the 3D MOC/DD method and a parallelization technique to apply the 3D MOC/DD method to a whole-core problem. The accuracy and performance are examined by solving the C5G7 benchmark and the BEAVRS whole-core benchmark problems

Adjoint flux calculation of natural mode equation by time dependent neutron transport

ABSTRACT The physical implication of the eigenfunction of the adjoint of the natural mode equation is studied. The eigenfunction at a phase space position is theoretically derived to be proportional to the amplitude of the neutron flux sufficiently long time after a source is placed at the position. Meanwhile, the time change rate and distribution of the neutron flux originated in the source must converge to those of the fundamental mode of the natural mode equation. Based on the proportional relation, the adjoint flux is calculated for a MOX- and UO2-fueled lattice by a continuous-energy Monte Carlo time-dependent neutron transport calculation. The calculated distributions of the adjoint flux agree well with those of the iterated fission probability both in the prompt and delayed critical states.

Activation of high flux test module sample holder after IFMIF-DONES operation

Nuclear safety assessment in nuclear fusion devices relies on the Monte Carlo method based neutron transport calculations. This paper presents information about the calculation results of the activities and dose rates caused by neuron irradiation for the structural materials of the high flux test module sample holder of IFMIF-DONES. The neutron induced activities and dose rates at shutdown were calculated by means of the FISPACT-2010 code with data from the EAF-2010 nuclear data library. Neutron fluxes and spectra were obtained with MCNP neutron transport calculations. The activities and dose rates were calculated at the end of irradiation of the assumed device operation scenario for cooling times of 0 s – 1000 year. In addition, radionuclides with contribution of at least 0.5% to the total value of activation characteristics at the previously mentioned cooling times were identified. After the operation, the most active radionuclide is 55Fe, with an activity share ranging from 30% (M200) to 63% (M8), and at the end of the prediction it accounts for 86% of the total activity. The highest dose rates at the end of irradiation are attributed to 56Mn radionuclide. 54Mn and 60Co are the most dominant radionuclides during intermediate and long cool-down periods.

Shielding analysis for TCP and IVVS bio-shield plugs of ITER

ITER neutronics model (Tokamak Complex) keeps evolving, and each development requires comprehensive transport analysis for verification and validation. The tokamak complex’s bio-shield plugs are the main shielding barriers between the tokamak and the port cells. Therefore, shielding analysis and optimization for the bio-shield plugs are part of the key factors for ITER licensing. In this work, performance evaluation of two designed bio-shield plugs: IVVS-port cell #3 and TCP-port cell #4 in level B1 has been conducted. Neutron transport calculation analysis on the new designs (generic designs) showed total neutron flux of 1.71 × 106 n/cm2/s and 2.86 × 105 n/cm2/s respectively in a specified tally cell behind the IVVS and TCP bio-shield plug. Spatial neutron distribution behind the bios-shield plug (port cells) showed significant shielding improvement compared to the original IVVS and TCP bio-shield design. Further iterative optimized shielding scheme has been proposed to reduce the total neutron flux from 1.71 × 106 to 8.8 × 104 n/cm2/s and from 2.86 × 105 to 6.4 × 104 n/cm2/s for the respective tally cell behind IVVS and TCP bio-shield plugs.

A Highly Parallel Multilevel Newton--Krylov--Schwarz Method with Subspace-Based Coarsening and Partition-Based Balancing for the Multigroup Neutron Transport Equation on Three-Dimensional Unstructured Meshes

The multigroup neutron transport equation is crucial for studying the motion of neutrons and their interaction with materials. Numerical simulation of the multigroup neutron transport equation is computationally challenging because the equation is defined on a high-dimensional phase space, the computational spatial domain is complex, and the materials are heterogeneous. A scalable parallel solver is required to address such a challenge. In this paper, we study a highly parallel Newton--Krylov--Schwarz (NKS) method consisting of a Newton-based eigenvalue solver, a Krylov subspace method, and a novel multilevel Schwarz preconditioner. The multilevel method is one of the most popular preconditioners for accelerating neutron transport calculations, but the construction of coarse spaces can be expensive and often unscalable when a large number of processors is used. We propose a novel matrix coarsening algorithm in which a multilevel hierarchy is constructed using a single-component matrix instead of the full matrix of the neutron transport equation. This new coarsening algorithm is referred to as “subspace-based coarsening.” Above 8,000 processors, we show a 13x enhancement in multilevel preconditioner setup time when using the subspace-based coarsening method. A partition-based balancing strategy is studied to enhance the parallel efficiency of the NKS algorithm by equalizing the work for each processor. A hierarchical mesh partitioning algorithm is employed to generate a large number of submeshes while minimizing off-node communication. We demonstrate that the proposed algorithm is scalable with more than 10,000 processors for a realistic application on three-dimensional unstructured meshes with a few billion degrees of freedom. Neutron transport calculations using the improved NKS algorithm are twice as fast as those based on the unmodified NKS solver when over 8,000 processors are employed.

Benchmarking of MANCINTAP 2.0, a parallel high-performance tool for neutron activation analysis in complex 4D scenarios

MANCINTAP is a parallel computational tool developed and maintained by Ansaldo Nucleare since 2011 to perform 4D neutron transport, activation and time-resolved dose-rate calculations in complex geometries for CPU-intensive applications. MANCINTAP creates an automated link between the 3D radiation transport code MCNP5—which is used to evaluate both the neutron fluxes for activation calculations and the resulting secondary photon dose rates—and the zero-dimensional activation code Anita2000 by handling crucial processes such as data exchange, determination of material mixtures and generation of cumulative probability distributions. Ansaldo Nucleare recently developed a new version of the tool, named MANCINTAP 2.0. A brief description of the computational tool is given here, with particular emphasis on the key technical choices underlying the project. The main enhancement is here highlighted with respect to the former version, and the results of a benchmarking campaign of MANCINTAP 2.0 against the Monte Carlo code FLUKA are described for two test cases.

New evaluation of general purpose neutron data for stable W-isotopes up to 200 MeV

In the frame of the Power Plant Physics and Technology of EUROfusion, new evaluations of general purpose neutron cross-section data were performed for the 180,182,183,184,186W isotopes covering the neutron energy up to 200 MeV. A special version of the TALYS nuclear model code implementing the geometry dependent hybrid model supplied with models for the non-equilibrium cluster emission was applied for calculations of the nuclide production and the energy distribution of the emitted particles. The parameters of the GDH model were properly estimated using measured data for individual tungsten isotopes. The neutron cross-sections were evaluated making use of available experimental data, systematics including estimated A-dependence of components of gas production cross-sections, and covariance information produced as part of the evaluation process. The BEKED code package, developed at KIT, was applied for calculations of co-variances using a dedicated Monte Carlo method. The evaluated data were processed into standard ENDF data format using the TEFAL code and the FOX module of the BEKED system. The evaluated data files were checked for errors and inconsistencies, processed with the NJOY code into ACE data format, and benchmarked against available integral experiments with MCNP neutron transport calculations.

Dose rates and specific activities of copper based materials irradiated during the TT plasma operation at JET

Abstract Before deuterium-tritium (DTE2), a new tritium campaign (TT) with the objective to validate the activities, dose rates for selected materials and OLTIS container components, as resulting from neutron irradiation at Joint European Torus (JET), is planned. This paper describes the activation and dose rate calculation performed for copper based materials irradiated throughout the foreseen TT plasma operation. Preparatory experiments and calculations are in progress for TT campaign to measure neutron flux spectrum at the irradiation position using dosimetry foils. In addition, the materials to be considered for irradiation during the TT campaign of the OLTIS were identified and analyzed in this study: Al-Bronze and CuCrZr. The neutron-induced activities and dose rates at shutdown were calculated by activation inventory code FISPACT-2010 using the neutron fluxes and spectra provided by the detailed MCNP5 neutron transport calculations. After the irradiation, the activities and dose rates are calculated at several cooling times after the end of irradiation. Highest dose rates the end of irradiation are attributed to Cu-64 radionuclide. Co-58 and Co-60 are the most dominant radionuclides during intermediate and long cooling periods. Highest activities are ascribed to Cu-64 at the beginning and Ni-63 at the end of investigation time.

Optimal time step length and statistics in Monte Carlo burnup simulations

Monte Carlo burnup simulations continue to be seen as computationally very expensive numerical routines despite recent developments of associated methods. Here, we suggest a way of improving the computing efficiency via optimisation of the length of the time steps and the number of neutron histories that are simulated at each Monte Carlo criticality run. So far, users of Monte Carlo burnup codes have been required to set these parameters at will; however, an inadequate choice of these free parameters can severely worsen the computing efficiency. We have tested a large number of combinations of the free parameters on a simplified and fast solver, and we have observed that the computing efficiency was maximized when the computing cost of all Monte Carlo neutron transport calculations (summed over all time steps) was approximately comparable to costs of other procedures (all depletion simulations, the loading and processing of neutron cross sections, etc.). In this technical note, we demonstrate these results, and we also derive a simple theoretical model of the convergence of Monte Carlo burnup simulations that conforms to these numerical results. Here, we also suggest a straightforward way to automatise the selection of the optimal values of the free parameters for Monte Carlo burnup simulations.

A parallel DSA algorithm in the 3-D cylindrical geometry

A parallel DSA algorithm based on the red-black sweep PCG method is proposed to accelerate the neutron transport calculation in the 3-D cylindrical geometry. This parallel acceleration technique shares the same spatial domain decomposition strategy with the KBA method. The red-black sweep iteration is formed by moving the r- and θ-axial components of the diffusion operator into the right hand side of the diffusion equation. Using the initial values refreshed by several red-black sweep iterations simplifies the matrix-vector multiplication in the CG procedure. Through the homogeneous problems, how the times of the red-black sweep effects the DSA algorithm is analyzed. A wide range of problems including FBR benchmark, PWR Pin-by-pin calculation and PWR shielding calculation have been performed to test the effectiveness and efficiency of the new method. The solution time is reduced by 3 ~ 10 times in different problems. This new acceleration algorithm has well parallel scalability.

Comparison of the performance of heterogeneous variational nodal method and 2D/1D-MOC on pin-resolved neutron transport calculations of PWR

Heterogeneous Variational Nodal Method (HVNM) and the 2D/1D Method of Characteristics (2D/1D-MOC) are two promising methods for high-fidelity neutronics calculation. HVNM starts from even-parity form of neutron transport equation and employs iso-parameteric finite elements, orthogonal polynomials and piece-wise constants for spatial discretization. Additionally, spherical harmonics are applied for the angular expansion. In terms of 2D/1D-MOC, it starts from first-order transport equation and splits the three-dimensional transport equation into one 2D and one 1D equations which are coupled by the radial and axial leakage terms. Then space and angle are discretized by characteristic lines in radial plane and 1D Sn calculation is employed in axial direction. Both HVNM and 2D/1D-MOC have the capability of dealing with pin-resolved geometry and strong transport effect in PWR core problem. Thus, we developed two codes PANX and NECP-X which based on the methods of HVNM and MOC respectively. In this paper, to compare their performance on pin-resolved problems in PWR, the KAIST problem, the NuScale problem and the Beavrs problem are calculated. Both computational accuracy and efficiency are evaluated. The numerical results shown that both PANX and 2D/1D-MOC can obtain accurate eigenvalues with the error less than 100 pcm. However, PANX obtains more accurate pin power distribution and achieves higher computational efficiency than 2D/1D-MOC.

Radiation Dose Analysis in Criticality Accident of Fuel Debris in Water

Removal of fuel debris is regarded as one of the most important operations in the decommissioning of the Fukushima Daiichi nuclear power station (1F-NPS) to decrease long-term risk. To begin the operation, the consequences of possible criticality accidents must be evaluated in advance. In this work, we evaluated radiation doses during possible criticality accidents at 1F-NPS in assumptive fuel debris systems. In particular, the relationship between the water level surrounding the fuel debris and the radiation dose was investigated. This is because the water level surrounding the fuel debris is thought to have an impact on radiation dose during accidents as it affects both the reactivity and shielding of radiation. A combination of space-dependent kinetic analysis and radiation transport analysis was carried out in order to consider the special characteristics of fuel debris systems in water. Instead of traditional point-kinetics analysis, we used the Multi-region Integral Kinetic (MIK) code, which is a unique method based on Monte Carlo neutron transport calculations. The radiation transport calculation code Particle and Heavy Ion Transport Code System (PHITS) was used as well. The analyses revealed that the dose caused by criticality accidents may be the largest in systems in which part of the fuel debris is exposed to the air.

Po-210 production in the European DEMO fusion reactor

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.

Implementation of the CPU/GPU hybrid parallel method of characteristics neutron transport calculation using the heterogeneous cluster with dynamic workload assignment

In recent years, graphics processing units (GPUs) have been adopted in many High-Performance Computing (HPC) systems due to their massive computational power and superior energy efficiency. And accelerating CPU-version computational code on heterogeneous clusters with multi-core CPUs and GPUs has attracted a lot of attention. One of the focus on heterogeneous computing is to efficiently take advantage of all computational resources, including both CPU and GPU available on a cluster. In this paper, a heterogeneous MPI + OpenMP/CUDA parallel algorithm for solving the 2D neutron transport equation with the method of characteristic (MOC) is implemented. In this algorithm, the spatial domain decomposition technique provides the coarse-grained parallelism with the MPI protocol while the fine-grained parallelism is exploited through OpenMP (in CPU calculated domain) and CUDA (in GPU calculated domain) based on the ray parallelization. In order to efficiently leverage the computing power of heterogeneous clusters, a dynamic workload assignment scheme is proposed, which is to distribute the workload based on the runtime performance of CPUs and GPUs in the cluster. Moreover, the strong scaling performance of the MPI + CUDA parallelization is studied through a performance analysis model which provides the detailed impact of the degradation in iteration scheme, the load imbalance issue, the data copy between CPUs and GPUs, and the MPI communication in the MPI + CUDA parallel algorithm. And the corresponding conclusion is still tenable for the MPI + OpenMP/CUDA parallelization. The C5G7 2D benchmark and an extended 2D whole-core problem are calculated with MPI + CUDA parallelization, MPI + OpenMP/CUDA parallelization, and the MPI parallelization for comparison. Numerical results demonstrate that the heterogeneous parallel algorithm maintains the desired accuracy. And the dynamic workload assignment scheme can provide the optimal workload assignment which ideally matches the experimental results. In addition, over 11% improvement is observed in MPI + OpenMP/CUDA parallelization compared against the MPI + CUDA parallelization. Moreover, the CPUs/GPUs heterogeneous clusters significantly outperform the CPUs clusters and one heterogeneous node shows basically five times faster than a CPUs node.

Multiphysics approach to plasma neutron source modelling at the JET tokamak

A novel multiphysics methodology for the computation of realistic plasma neutron sources has been developed. The method is based on state-of-the-art plasma transport and neutron spectrum calculations, coupled with a Monte Carlo neutron transport code, bridging the gap between plasma physics and neutronics. In the paper two JET neutronics tokamak models are used to demonstrate the application of the developed plasma neutron sources and validate them. Diagnostic data for the record JET D discharge 92436 are used as input for the TRANSP code, modelling neutron emission in two external plasma heating scenarios, namely using only neutral beam injection and a combination of the latter and ion cyclotron resonance heating. Neutron spectra, based on plasma transport results, are computed using the DRESS code. The developed PLANET code package is employed to generate plasma neutron source descriptions and couple them with the MCNP code. The effects of using the developed sources in neutron transport calculations on the response of JET neutron diagnostic systems is studied and compared to the results obtained with a generic plasma neutron source. It is shown that, although there are significant differences in the emissivity profiles, spectra shape and anisotropy between the neutron sources, the integral response of the time-resolved ex-vessel neutron detectors is largely insensitive to source changes, with major relative deviations of up to several percent. However it is calculated that, due to the broadening of neutron spectra as a consequence of external plasma heating, larger differences may occur in activation of materials which have threshold reactions located at DD neutron peak energies. The PLANET plasma neutron source computational methodology is demonstrated to be suitable for detailed neutron source effect studies on JET during DT experiments and can be applied to ITER analyses.

СРАВНЕНИЕ ГОМОГЕННОГО И ГЕТЕРОГЕННОГО ИСПОЛЬЗОВАНИЯ ЭНЕРГЕТИЧЕСКОГО ПЛУТОНИЯ В ВВЭР

There are separate of fissile isotopes in a fuel rod in the fuel load admit the transition to a closed fuel circle in the nuclear power industry. For this solution to consider the formation of heterogeneous fuel pellet in which the fissile and raw isotope are separate and not mixed together. The raw material isotope is located in the peripheral zone. Fissile isotope, placed in the center of fuel pin holds 10% of the volume. There is row isotope in the periphery. All calculations for this research were made with UNK code. This code was complemented with a module, which allows calculating temperatures in pin cell. At present UNK allows solving of neutron-physical task of neutron transport, burn-up task (isotope composition evolution) and temperature calculation. Initial loading is plutonium, separated from spent fuel of the same cell of previous company. Irradiated plutonium of the central part is sent to either long term storage or final disposal. Energy plutonium location extracted from spent fuel of VVER in the central part of fuel pellet (heterogeneous pellet) decreases plutonium loading approximately 20 % in condition of the same company duration in comparison with UOX fuel. In case of heterogeneous location, there is depleted uranium in the periphery on the pellet. And after irradiation received plutonium isotope composition can be used repeatedly to form new loadings. This solution is qualitatively different from MOX case, which does not allow further using.

Валидация пакета программ MCPV, адаптированного для расчетов переноса нейтронов в околокорпусном пространстве реактора ВВЭР-440

Валидация пакета программ MCPV, адаптированного для расчетов переноса нейтронов в околокорпусном пространстве реактора ВВЭР-440

GPU accelerated lattice Boltzmann method in neutron kinetics problems II: Neutron transport calculation

The simulation of neutron transport equation (NTE) has become one of the main trends of nuclear reactor engineering and design owing to its high accuracy, wherein one of its limitations is the low computational speed. To improve this condition, this paper provides a GPU accelerated neutron transport lattice Boltzmann method (NTLBM) for solving neutron transport problem. The simplified finite Boltzmann scheme and the GPU CUDA technique are combined to achieve the spatial parallel calculation and the space-angle full parallel calculation. Due to the ease of implementation and high parallelism, the GPU accelerated NTLBM (GPU-NTLBM) can be very easily and efficiently to realize. Besides, by using the multi-GPU acceleration, the large-scale NTLBM engineering calculation can be efficiently speeded up, and the properties of NTLBM makes the internal boundary treatments convenient. In considering the properties of the NTE, the space-angle full parallel acceleration is applied to further speed up the GPU-NTLBM for the first time, in which all the spatial and the angular dimensions are calculated in parallel. Simulation results show that the NTLBM can be applied to solve the multi-group neutron transport problems accurately and the GPU acceleration can be applied to effectively speed up the NTLBM calculation. The space-angle full parallel acceleration can further improve the GPU-NTLBM acceleration. This paper could provide some parallel perspectives for solving the large-scale neutron transport engineering calculation.