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Author: Publisher: ISBN: Category : Languages : en Pages :
Book Description
A numerical method is presented for solving the time-dependent survival probability equation in general (lD/2D/3D) geometries using the multi group SNmethod. Although this equation was first formulated by Bell in the early 1960's, it has only been applied to stationary systems (for other than idealized point models) until recently, and detailed descriptions of numerical solution techniques are lacking in the literature. This paper presents such a description and applies it to a dynamic system representative of a figurative criticality accident scenario.
Author: Publisher: ISBN: Category : Languages : en Pages :
Book Description
A numerical method is presented for solving the time-dependent survival probability equation in general (lD/2D/3D) geometries using the multi group SNmethod. Although this equation was first formulated by Bell in the early 1960's, it has only been applied to stationary systems (for other than idealized point models) until recently, and detailed descriptions of numerical solution techniques are lacking in the literature. This paper presents such a description and applies it to a dynamic system representative of a figurative criticality accident scenario.
Author: Publisher: ISBN: Category : Languages : en Pages :
Book Description
A new deterministic-Monte Carlo hybrid solution technique is derived for the time-dependent transport equation. This new approach is based on dividing the time domain into a number of coarse intervals and expanding the transport solution in a series of polynomials within each interval. The solutions within each interval can be represented in terms of arbitrary source terms by using precomputed response functions. In the current work, the time-dependent response function computations are performed using the Monte Carlo method, while the global time-step march is performed deterministically. This work extends previous work by coupling the time-dependent expansions to space- and angle-dependent expansions to fully characterize the 1D transport response/solution. More generally, this approach represents and incremental extension of the steady-state coarse-mesh transport method that is based on global-local decompositions of large neutron transport problems. An example of a homogeneous slab is discussed as an example of the new developments.
Author: James Ernest Banfield Publisher: ISBN: Category : Neutrons Languages : en Pages : 170
Book Description
Using a semi-implicit direct kinetics (SIDK) method that is developed in this dissertation, a finer neutron energy discretization and improved fidelity for transient radiation transport calculations are facilitated to reduce uncertainties and conservatisms in transient power and temperature predictions. These capabilities are implemented within a parallel computational solver framework, which is able to represent an arbitrary number of neutron energy groups, angles, and spatial discretizations, while internally coupled to an unstructured finite element multi-physics code for temperature and displacement calculations. This capability is demonstrated on a three-dimensional control rod ejection simulation run in parallel utilizing forty-four neutron energy groups. An improved transient nuclear reactor simulation capability is developed by adapting the steady-state radiation transport code Denovo to solve the time-dependent Boltzmann transport equation for transient power distributions. The developed SIDK method is compared to fully-implicit direct kinetics, higher order time integration methods, as well as various computational benchmarks. Errors resulting from time integration, spatial discretization, angular treatment, multi-group treatment, homogenization of temperature, and power over the time step representation are explored. For verification, the SIDK method is developed and tested externally and independently employing a few-group time-dependent neutron diffusion code which is compared to one and two-dimensional benchmarks with and without temperature feedbacks. The results of the semi-implicit direct kinetics method (SIDK) are shown to be accurate to within ~0.2% of direct kinetics and to execute roughly an order of magnitude faster, using a consistent space and time discretization. For sufficiently severe transients, the direct method is shown to produce lower errors with medium time steps than the SIDK method with fine steps, but proves to be subject to more severe oscillations at very coarse time steps than the SIDK method, in addition to producing similar errors (within 0.2 %) at medium spatial discretization with consistent time steps. The objective of this dissertation is to provide developers of next generation high-performance computing neutron kinetics methods a guide to the benefits and costs of the dominant discretization strategies of time, space, neutron energy, and angle for the solution of the time-dependent Boltzmann transport equation.
Author: Publisher: ISBN: Category : Languages : en Pages : 8
Book Description
A deterministic method for solving the time-dependent, three-dimensional Boltzmann transport equation with explicit representation of delayed neutrons has been developed and evaluated. The methodology used in this study for the time variable of the neutron flux is known as the improved quasi-static (IQS) method. The position, energy, and angle-dependent neutron flux is computed deterministically by using the three-dimensional discrete ordinates code TORT. This paper briefly describes the methodology and selected results. The code developed at the University of Tennessee based on this methodology is called TDTORT. TDTORT can be used to model transients involving voided and/or strongly absorbing regions that require transport theory for accuracy. This code can also be used to model either small high-leakage systems, such as space reactors, or asymmetric control rod movements. TDTORT can model step, ramp, step followed by another step, and step followed by ramp type perturbations. It can also model columnwise rod movement. A special case of columnwise rod movement in a three-dimensional model of a boiling water reactor (BWR) with simple adiabatic feedback is also included. TDTORT is verified through several transient one-dimensional, two-dimensional, and three-dimensional benchmark problems. The results show that the transport methodology and corresponding code developed in this work have sufficient accuracy and speed for computing the dynamic behavior of complex multi-dimensional neutronic systems.
Author: Guillaume Louis Giudicelli Publisher: ISBN: Category : Languages : en Pages : 542
Book Description
With ever increasing available computing resources, the traditional nuclear reactor physics computation schemes, that trade off between spatial, angular and energy resolution to achieve low cost highly-tuned simulations, are being challenged. While existing schemes can reach few-percent accuracy for the current fleet of light water reactors, thanks to a plethora of astute engineering approximations, they cannot provide sufficient accuracy for evolutionary reactor designs with highly heterogeneous geometries. The decades-long process to develop and qualify these simulation tools is also not in phase with the fast-paced development of innovative new reactor designs seeking to address the climate crisis. Enabled by those computing resources, high fidelity Monte Carlo methods can easily tackle challenging geometries, but they lack the computational and algorithmic efficiency of deterministic methods. However, they are increasingly being used for group cross section generation. Downstream highly parallelized 3D deterministic transport can then use those cross sections to compute accurate solutions at the full core scale. This hybrid computation scheme makes the most of both worlds to achieve fast and accurate reactor physics simulations. Among the few remaining approximations are neglecting the angular dependence of group cross sections, which lead to an over-estimation of resonant absorption rates, especially for the lower resonances of 238U. This thesis presents a novel equivalence method based on introducing discontinuities in the track angular fluxes, with a polar dependence of discontinuity factors to preserve the polar dependence of the neutron currents as well as removing the self-shielding error. This new method is systematically benchmarked against the state-of-the-art method, SuPerHomogenization in three different approaches to obtaining equivalence factors: a same-scale iterative approach, a multiscale approach, and a single-step non-iterative approach. Both methods show remarkable agreement with a reference Monte Carlo solution on a wide array of test cases, from 2D pin cells to 3D full core calculations, for the iterative and the multi-scale approaches. The self-shielding error is eliminated, improving significantly the predictive capabilities of the scheme for the distribution of 238U absorption in the core. A single-step non-iterative approach to obtaining equivalence factors is also pursued, and was shown to only be adequate with the novel discontinuity factor-based method. This study is largely enabled by a significant optimization effort of the 3D deterministic neutron transport solver. By leveraging low level parallelism through vectorization of the multi-group neutron transport equation, by increasing the memory locality of the method of characteristics implementation and with a novel inter-domain communication algorithm enabling a near halving of memory requirements, the 3D full core case can now be tackled with only 50 nodes on an industrial sized computing cluster rather than the many thousands of nodes on a TOP20 supercomputer used previously. This thesis presents fully resolved solutions to the steady-state multi-group neutron transport equation for full-core 3D light water reactors, and these solutions are comparable to gold-standard continuous-energy Monte Carlo solutions.
Author: Publisher: ISBN: Category : Languages : en Pages :
Book Description
A deterministic method for solving the time-dependent, three-dimensional Boltzmam transport equation with explicit representation of delayed neutrons has been developed and evaluated. The methodology used in this study for the time variable of the neutron flux is known as the improved quasi-static (IQS) method. The position, energy, and angle-dependent neutron flux is computed deterministically by using the three-dimensional discrete ordinates code TORT. This paper briefly describes the methodology and selected results. The code developed at the University of Tennessee based on this methodology is called TDTORT. TDTORT can be used to model transients involving voided and/or strongly absorbing regions that require transport theory for accuracy. This code can also be used to model either small high-leakage systems, such as space reactors, or asymmetric control rod movements. TDTORT can model step, ramp, step followed by another step, and step followed by ramp type perturbations. It can also model columnwise rod movement can also be modeled. A special case of columnwise rod movement in a three-dimensional model of a boiling water reactor (BWR) with simple adiabatic feedback is also included. TDTORT is verified through several transient one-dimensional, two-dimensional, and three-dimensional benchmark problems. The results show that the transport methodology and corresponding code developed in this work have sufficient accuracy and speed for computing the dynamic behavior of complex multidimensional neutronic systems.
Author: Christopher James Hurt Publisher: ISBN: Category : Isotopes Languages : en Pages : 222
Book Description
The computational ability to accurately predict the conditions in an experiment under irradiation is a valuable tool in the operation of a research reactor whose scientific mission includes isotope production, materials irradiation, and neutron activation analysis. Understanding of different governing physics is required to ascertain satisfactory conditions within the experiment: the neutron transport behavior throughout the reactor and the coupled behavior of heat transfer, structural mechanics and fluid flow. Computational methods and tools were developed for robust numerical analysis of experiment behavior at the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR), including fully-coupled thermo-mechanics in three plutonium-238 (238Pu) production targets. In addition, a new computational tool was developed that solves neutron transport using the discrete ordinates method on a finite element mesh and offers multiphysics coupling. The thermo-mechanical models of the 238Pu targets are solved using the COMSOL heat transfer and solid mechanics modules with irradiation behavior and thermophysical properties taken from measurements performed at ORNL. The experiments, placed in the permanent beryllium (PB) reflector, consist of neptunium dioxide/aluminum (NpO2/Al) pellets in Al containment, the model taking advantage of axisymmetry in two-dimensional R-Z cylindrical geometry. At times, extended analysis was needed for incomplete data sets and time schedules; however, the thermal-structure models ensured progression through three project phases of target qualification. The neutron transport equation was solved in COMSOL, using the discrete ordinates formulation in the weak form partial differential equation (PDE) interface. Validation studies were performed for the dimensions developed (one-, two- and three- dimensional Cartesian as well as two-dimensional R-Z cylindrical/axi-symmetric) and compared to external deterministic and stochastic codes. The method was then applied to a beginning-of-cycle (BOC) simplified HFIR core, with good comparison to other static solutions of the HFIR, and a time-dependent extension to this tool was created and exhibited for a benchmark problem. The research presented in this dissertation is the continued progress towards creating a comprehensive multiphysics methodology for studying the dynamic behavior of the HFIR core, and shows the capabilities of detailed space-time reactor physics studies and of multiphysics analyses for experiment qualification and safety analyses at a research reactor.
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Author: James Ernest Banfield
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Author: Guillaume Louis Giudicelli
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Author: Bengt G. Carlson
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Author: Christopher James Hurt