1,721,088 research outputs found
Solving eigenvalue response matrix equations with nonlinear techniques
Full text linkThis paper presents new algorithms for use in the eigenvalue response matrix method (ERMM) for reactor eigenvalue problems. ERMM spatially decomposes a domain into independent nodes linked via boundary conditions approximated as truncated orthogonal expansions, the coefficients of which are response functions. In its simplest form, ERMM consists of a two-level eigenproblem: an outer Picard iteration updates the k-eigenvalue via balance, while the inner λ -eigenproblem imposes neutron balance between nodes. Efficient methods are developed for solving the inner λ-eigenvalue problem within the outer Picard iteration. Based on results from several diffusion and transport benchmark models, it was found that the Krylov-Schur method applied to the λ -eigenvalue problem reduces Picard solver times (excluding response generation) by a factor of 2–5. Furthermore, alternative methods, including Picard acceleration schemes, Steffensen’s method, and Newton’s method, are developed in this paper. These approaches often yield faster k-convergence and a need for fewer k-dependent response function evaluations, which is important because response generation is often the primary cost for problems using responses computed online (i.e., not from a precomputed database). Accelerated Picard iteration was found to reduce total computational times by 2–3 compared to the unaccelerated case for problems dominated by response generation. In addition, Newton’s method was found to provide nearly the same performance with improved robustness
Bond graph representation of nuclear reactor point kinetics and nearly incompressible thermal hydraulics
No full textThis work presents a simplified 1D model for a pressurized water reactor core, suitable for very rapid transients like control rod ejection. The model is represented using the bond graph formalism, a technique for modeling engineering systems as combinations of connected elements. Bond graphs are a flexible way of presenting coupled physics problems by automating the computer science aspects of modeling and letting the modelers focus on the physics; they were introduced in earlier work. To help leverage the flexibility of bond graph representations of physical systems, a new bond graph processing code, BGSolver, is introduced. BGSolver has been developed by the authors over the past several years, and is now released as open source software. A rapid rod ejection benchmark is solved with both BGSolver and RELAP5-3D; BGSolver obtained full convergence with a 5 ms time step, while RELAP5-3D required a 1 ms time step, due to the fully coupled time integration that BGSolver employed, compared to an operator splitting-based time integrator of RELAP5-3D. BGSolver’s time integrator demonstrated 3rd-order convergence in time, a very desirable property. A single nonlinear solve was used to obtain the steady state with BGSolver.United States. Office of the Assistant Secretary for Nuclear Energy (Nuclear Energy University Program fellowship
A Generalized Optimization Methodology for Isotope Management
No full textThis research focuses on developing a new approach to studying the nuclear fuel cycle:
instead of employing the trial and error approach currently used in actinide management
studies in which reactors are designed and then their performance is evaluated, the
methodology developed here first identifies relevant fuel cycle objectives–like minimizing
decay heat production in a repository, minimizing Pu-239 content in used fuel, etc.–and then
uses optimization to determine the best way to reach these goals.
The first half of this research was devoted to identifying optimal flux spectra for irradiating
used nuclear fuel from light water reactors to meet fuel cycle objectives like those mentioned
above. This was accomplished by applying the simulated annealing optimization
methodology to a simple matrix exponential depletion code written in Fortran using cross
sections generated from the SCALE code system.
Since flux spectra cannot be shaped arbitrarily, the second half of this research applied the
same methodology to material composition of fast reactor target assemblies to find optimal
designs for minimizing the integrated decay heat production over various timescales. The
neutronics calculations were performed using modules from SCALE and ERANOS, a French
fast reactor transport code.United States. Dept. of Energy. Office of Nuclear Energy (Fuel Cycle Research and Development Program
General Analysis of Breed-and-Burn Reactors and Limited-Separations Fuel Cycles
No full textA new theoretical framework is introduced, the “neutron excess” concept, which is useful
for analyzing breed-and-burn (B&B) reactors and their fuel cycles. Based on this concept, a
set of methods has been developed which allows a broad comparison of B&B reactors using
different fuels, structural materials, and coolants. This new approach allows important
reactor and fuel-cycle parameters to be approximated quickly, without the need for a full
core design, including minimum burnup/irradiation damage and reactor fleet doubling time.
Two general configurations of B&B reactors are considered: a “minimum-burnup” version
in which fuel elements can be shuffled in three dimensions, and a “linear-assembly” version
composed of conventional linear assemblies that are shuffled radially.
Based on studies of different core compositions, the best options for minimizing fuel burnup
and material DPA are metal fuel (with a strong dependence on alloy content), the type of
steel that allows the lowest structure volume fraction, and helium coolant. If sufficient fuel
performance margin exists, sodium coolant can be substituted in place of helium to achieve
higher power densities at a modest burnup and DPA penalty. For a minimum-burnup B&B
reactor, reasonably achievable minimum DPA values are on the order of 250-350 DPA in
steel, while axial peaking in a linear-assembly B&B reactor raises minimum DPA to over
450 DPA. By recycling used B&B fuel in a limited-separations (without full actinide
separations) fuel cycle, there is potential for sodium-cooled B&B reactors to achieve fleet
doubling times of less than one decade, although this result is highly sensitive to the reactor
core composition employed as well as thermal hydraulic performance.TerraPower, Inc.Massachusetts Institute of Technology. Energy Initiative (The Future of the Nuclear Fuel Cycle Study
Feasibility of Breeding in Hard Spectrum Boiling Water Reactors with Oxide and Nitride Fuels
No full textThis study assesses the neutronic, thermal-hydraulic, and fuel performance aspects of using
nitride fuel in place of oxides in Pu-based high conversion light water reactor designs. Using
the higher density nitride fuel hardens the neutron energy spectrum and results in higher
breeding ratios.
The state-of-the-art high conversion light water reactor, the Resource-renewable
Boiling Water Reactor (RBWR), served as the template core upon which comparative studies
between nitride and oxide fuels were performed. A 1/3 core reactor physics model was
developed for the RBWR using the stochastic transport code MCNP. The code was coupled
with a lumped channel thermal-hydraulics 5-channel model for steady-state analyses. The
depletion code MCODE, which links MCNP with ORIGEN, was used for all burnup
calculations. Select physics parameters were calculated and with the exception of the void
coefficients, agreed with reported data. The void coefficients of the coupled core were
calculated to be slightly positive using two different methods (10% power increase and 5%
flow reduction).
The standard RBWR assembly designs, which use tight lattice hexagonal fuel rod
arrays, with oxide fuel were then replaced with various nitride fuel assembly designs to
determine the potential increase in breeding ratio, the potential to breed with pressurized water,
and the potential to improve the critical power ratio with a wider pin pitch. Without changing
the assembly geometry or discharge burnup, using nitride fuel resulted in a breeding ratio of
1.14. Using single-phase liquid water, the nitride fuel RBWR assembly resulted in a conversion
ratio of 1.00. Another nitride fuel assembly design with boiling water maintained a 1.04
breeding ratio while increasing the pitch-to-diameter ratio from 1.13 to 1.20. This modification
increased the hot assembly critical power ratio from 1.22 to 1.36, as calculated using the Liu-
2007 correlation.
A high-porosity nitride fuel is recommended for high burnup conditions, to
accommodate the nitride fuel’s higher swelling and less favorable mechanical properties
compared to the oxide fuel. The high porosity allows additional volume for pressure-induced
densification, alleviating swelling and subsequent cladding strain. To predict the performance
of high-porosity nitride fuel, fission gas and fuel behavior mechanistic models were developed
for high burnup and low-temperature conditions. These models were validated with reported
irradiation data and implemented, along with fuel material properties, into the steady-state fuel
behavior code FRAPCON-EP. Under simulated RBWR conditions, a fuel density no more than
85% of theoretical density is recommended to maintain satisfactory fuel performance.ArevaMassachusetts Institute of Technology Study on the Future on The Nuclear Fuel Cycl
Parallel Fission Bank Algorithms in Monte Carlo Criticality Calculations
Full text linkIn this work we describe a new method for parallelizing the source iterations in a Monte Carlo criticality calculation. Instead of having one global fission bank that needs to be synchronized, as is traditionally done, our method has each processor keep track of a local fission bank while still preserving reproducibility. In doing so, it is required to send only a limited set of fission bank sites between processors, thereby drastically reducing the total amount of data sent through the network. The algorithm was implemented in a simple Monte Carlo code and shown to scale up to hundreds of processors and furthermore outperforms traditional algorithms by at least two orders of magnitude in wall-clock time
An energy recondensation method using the discrete generalized multigroup energy expansion theory
Full text linkIn this paper, the discrete generalized multigroup (DGM) method was used to recondense the coarse group cross-sections using the core level solution, thus providing a correction for neighboring effect found at the core level. This approach was tested using a discrete ordinates implementation in both 1-D and 2-D. Results indicate that 2 or 3 iterations can substantially improve the flux and fission density errors associated with strong interfacial spectral changes as found in the presence of strong absorbers, reflector of mixed-oxide fuel. The methodology is also proven to be fully consistent with the multigroup methodology as long as a flat-flux approximation is used spatially
Accelerated sampling of the free gas resonance elastic scattering kernel
No full textIn this work, we present the derivation and investigation of a new Doppler broadening rejection sampling approach for the exact treatment of resonance elastic scattering in Monte Carlo neutron transport codes. Implemented in OpenMC, this method correctly accounts for the energy dependence of cross sections when treating the thermal motion of target nuclei in elastic scattering events. The method is verified against both stochastic and deterministic reference results in the literature for ²³⁸U resonance scattering. Upscatter percentages and mean scattered energies calculated with the method are shown to agree well with the reference scattering kernel results. Additionally, pin cell and full core k[subscript eff] results calculated with this implementation of the exact resonance scattering kernel are shown to be in close agreement with those in the literature. The attractiveness of the method stems from its improvement upon a computationally expensive rejection sampling procedure employed by an earlier stochastic resonance scattering treatment. With no loss in accuracy, the accelerated sampling algorithm is shown to reduce overall runtime by 3–5% relative to the Doppler broadening rejection correction method for both pin cell and full core benchmark problems. This translates to a 30–40% reduction in runtime overhead.United States. Department of Energy (DE-AC05-00OR22725
On the stability of the Discrete Generalized Multigroup method
No full textThis paper investigates the stability of the recondensation procedure of the Discrete Generalized Multigroup method and proposes alternatives to improve stability of the original formulation. Instabilities are shown to happen when employing a simple Picard fixed point iteration and an ill-informed group mapping scheme. This work presents a mapping procedure that improves stability of the original method for fine group calculations. Additionally, a relaxation scheme, Krasnoselskij iteration, is introduced to the fixed point iteration to further improve the stability characteristics and remove the need for fine group flux updates. Both improvements are applied on heterogeneous problems using the SHEM361 and the NG2042 group structures. The results indicate improved stability from a well-informed group mapping and demonstrate the possibility of eliminating the need for fine group flux updates.United States. Dept. of Energy. Naval Reactors Divisio
The OpenMC Monte Carlo particle transport code
Full text linkA new Monte Carlo code called OpenMC is currently under development at the Massachusetts Institute of Technology as a tool for simulation on high-performance computing platforms. Given that many legacy codes do not scale well on existing and future parallel computer architectures, OpenMC has been developed from scratch with a focus on high performance scalable algorithms as well as modern software design practices. The present work describes the methods used in the OpenMC code and demonstrates the performance and accuracy of the code on a variety of problems.United States. Department of Energy (DE-AC05-00OR22725
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