Multi-core performance studies of a Monte Carlo neutron transport code

Performance results are presented for a multi-threaded version of the OpenMC Monte Carlo neutronics code using OpenMP in the context of nuclear reactor criticality calculations. Our main interest is production computing, and thus we limit our approach to threading strategies that both require reasonable levels of development effort and preserve the code features necessary for robust application to real-world reactor problems. Several approaches are developed and the results compared on several multi-core platforms using a popular reactor physics benchmark. A broad range of performance studies are distilled into a simple, consistent picture of the empirical performance characteristics of reactor Monte Carlo algorithms on current multi-core architectures.United States. Dept. of Energy. Office of Advanced Scientific Computing Research (Contract DEAC02-06CH11357

Monte Carlo methodologies for neutron streaming in diffusion calculations - Application to directional diffusion coefficients and leakage models in XS generation

Neutron transport calculations by Monte Carlo methods are finding increased application in nuclear reactor simulations. In particular, a versatile approach entails the use of a 2-step procedure, with Monte Carlo as a few-group cross section data generator at lattice level, followed by deterministic multi-group diffusion calculations at core level. In this thesis, the Serpent 2 Monte Carlo reactor physics burnup calculation code is used in order to test a set of diffusion coefficient models, as well as neutron leakage methodologies at assembly level. The tests include novel anisotropic diffusion coefficient and heterogeneous leakage models developed and implemented by the author. The analyses are mainly focused on a sodium-cooled fast reactor system, for which few-group cross section data was generated by stochastic methods with Serpent 2. The quality of the full-core diffusion results is evaluated by contrasting system eigenvalues and power distributions against detailed, full-core reference solutions also supplied by the Serpent 2 code and the same nuclear data library. Whereas the new anisotropic diffusion coefficient formalism exhibits improved performance in the fast reactor system studied, there are restrictions to its applicability in other reactor designs. The newly proposed leakage model has a similar performance to that one of albedo ite-rations, and provides valuable information about the space-energy coupling of the scalar neutron flux at lattice level. This hitherto unavailable information does not entail a significant computational cost. In sodium-cooled fast reactor calculations, the quality of diffusion theory results can be improved by either using directional diffusion coefficients and a fine energy mesh, or via leakage-corrected discontinuity factors. These factors can be calculated using net neutron currents supplied by heterogeneous leakage models. Preliminary results from this research also suggest that the studies maybe extended to graphite-moderated, gas-cooled reactors

Improving PWR core simulations by Monte Carlo uncertainty analysis and Bayesian inference

A Monte Carlo-based Bayesian inference model is applied to the prediction of reactor operation parameters of a PWR nuclear power plant. In this non-perturbative framework, high-dimensional covariance information describing the uncertainty of microscopic nuclear data is combined with measured reactor operation data in order to provide statistically sound, well founded uncertainty estimates of integral parameters, such as the boron letdown curve and the burnup-dependent reactor power distribution. The performance of this methodology is assessed in a blind test approach, where we use measurements of a given reactor cycle to improve the prediction of the subsequent cycle. As it turns out, the resulting improvement of the prediction quality is impressive. In particular, the prediction uncertainty of the boron letdown curve, which is of utmost importance for the planning of the reactor cycle length, can be reduced by one order of magnitude by including the boron concentration measurement information of the previous cycle in the analysis. Additionally, we present first results of non-perturbative nuclear-data updating and show that predictions obtained with the updated libraries are consistent with those induced by Bayesian inference applied directly to the integral observables.Comment: 10 pages, 11 figure

Parallel performance results for the OpenMOC neutron transport code on multicore platforms

The shift toward multicore architectures has ushered in a new era of shared memory parallelism for scientific applications. This transition has introduced challenges for the nuclear engineering community, as it seeks to design high-fidelity full-core reactor physics simulation tools. This article describes the parallel transport sweep algorithm in the OpenMOC method of characteristics (MOC) neutron transport code for multicore platforms using OpenMP. Strong and weak scaling studies are performed for both Intel Xeon and IBM Blue Gene/Q (BG/Q) multicore processors. The results demonstrate 100% parallel efficiency for 12 threads on 12 cores on Intel Xeon platforms and over 90% parallel efficiency with 64 threads on 16 cores on the IBM BG/Q. These results illustrate the potential for hardware acceleration for MOC neutron transport on modern multicore and future many-core architectures. In addition, this work highlights the pitfalls of programming for multicore architectures, with a focal point on false sharing.National Science Foundation (U.S.). Graduate Research Fellowship Program (Grant 1122374)United States. Department of Energy (Center for Exascale Simulation of Advanced Reactors. Contract DE-AC02-06CH11357

Neutronics Studies on the NIST Reactor Using the GA LEU fuel

The National Bureau of Standards Reactor (NBSR) located on the National Institute of Standards and Technology (NIST) Gaithersburg campus, is currently underway of fuel conversion from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. One particular challenging part of the conversion of the NBSR is the high average flux level (2.5×1014 n/cm2-s) required to maintain experimental testing capabilities of the reactor, without significant changes to the external structures of the reactor. Recently the General Atomics (GA) Training Research Isotopes General Atomics (TRIGA) fuel has shown some promising features as a LEU candidate for the high performance research reactors such as the NBSR. The GA fuel has a long history of success in conversion of research reactors since it was developed in 1980s. The UZrH compound in the GA fuel has seen success in long term TRIGA reactors, and is a proven safe LEU alternative. This study performs a neutronics evaluation of the TRIGA fuel under the schema of the NBSR’s heavy conversion requirements in order to examine whether the TRIGA fuel is a viable option for conversion of the NBSR. To determine the most optimal path of conversion, we performed a feasibility study with particular regard to the fuel dimensions, fuel rod configurations, cladding, as well as fuel structure selection. Based on the outcome of the feasibility study, an equilibrium core is then generated following the NBSR’s current fuel management schema. Key neutronics performance characteristics including flux distribution, power distribution, control rod (i.e., shim arms) worth, as well as kinetics parameters of the equilibrium core are calculated and evaluated. MCNP6, a Monte Carlo based computational modeling software was intensively used to aid in these calculations. The results of this study will provide important insight on the effectiveness of conversion, as well as determine the viability of the conversion from HEU to LEU using the GA fuel

A Compact Beam Stop for a Rare Kaon Decay Experiment

We describe the development and testing of a novel beam stop for use in a rare kaon decay experiment at the Brookhaven AGS. The beam stop is located inside a dipole spectrometer magnet in close proximity to straw drift chambers and intercepts a high-intensity neutral hadron beam. The design process, involving both Monte Carlo simulations and beam tests of alternative beam-stop shielding arrangements, had the goal of minimizing the leakage of particles from the beam stop and the resulting hit rates in detectors, while preserving maximum acceptance for events of interest. The beam tests consisted of measurements of rates in drift chambers, scintilation counter hodoscopes, a gas threshold Cherenkov counter, and a lead glass array. Measurements were also made with a set of specialized detectors which were sensitive to low-energy neutrons, photons, and charged particles. Comparisons are made between these measurements and a detailed Monte Carlo simulation.Comment: 39 pages, 14 figures, submitted to Nuclear Instruments and Method

Modeling Neutron Interaction Inside a 2D Reactor Using Monte Carlo Method

Scientists and engineers have been working for many years to develop accurate approaches to analyzing nuclear power reactors using computer codes that closely model the behavior of neutrons in a reactor core. The Monte Carlo simulation method is capable of treating complex geometries with a high level of resolution and fidelity to model neutron interactions inside a reactor core. With the requirement of accurate modeling in reactor physics and dynamics and great innovation of computer technology, Monte Carlo method is becoming an ever more powerful tool and receiving rising attention. In this study, Monte Carlo method is used to model nuclear interactions between randomly moving neutrons and the fuel material, cladding material and moderator. The code, QualifyingMC, written using Python language develops the neutron diffusion scenario in a two-dimensional car- tesian geometry. To evaluate the performance and accuracy of the simulation, the calcu- lated values of the effective multiplication factor (keff), a key component in characterizing the breeding property of a fission-reactor system, was compared with reference values cal- culated with other codes using the same geometry, materials and boundary conditions. A good agreement within a few percent on multiplication factors was obtained. The neutron flux distribution, another important parameter in a fission-reactor system, as a function of neutron energy is also calculated and compared with the Watt distribution function. A rea- sonable agreement between QualifyingMC and the reference results was obtained