Advanced Burner Test Reactor Research Articles

Development of multi-objective core optimization framework and application to sodium-cooled fast test reactors

The optimization of a Sodium-cooled Fast Reactor (SFR) core is a challenging process, due to the large number of design parameters, the nonlinearities among inputs and outputs, and the complicated correlation among output parameters. This study attempts to develop a generalized framework for the SFR core optimization by coupling the sensitivity analysis, advanced optimization algorithm, and optionally the surrogate modeling. The framework is built based on the fast reactor modeling capability of the Argonne Reactor Computation (ARC) suite and the sensitivity analysis and optimization modules embedded in the DAKOTA code, both have been integrated within the NEAMS Workbench. The genetic algorithm is selected as the optimization method for its robustness, while the option of surrogate modeling is also explored to alleviate the computational burden caused by employing the ARC direct core physics simulation and thus enhance the efficiency of the optimization. Finally, the normalized deviations of performance parameters of the near-optimal solution from those of the ideal core are calculated and used as criteria to down select the final core design. The developed framework is applied to the Advanced Burner Test Reactor (ABTR) core, and optimal solutions are determined by balancing various objectives simultaneously, including peak fast flux, core volume, power, reactivity swing, plutonium mass feed, while at the same time satisfying the predefined constraints due to safety or economics considerations. The optimal ABTR core design obtained using the direct physical simulation and surrogated model are compared and discussed. It is found that using the accurately constructed surrogate models can significantly reduce the required computational time while maintaining satisfactory accuracy.

Coupled Neutron and Gamma Heating Calculation Based on VARIANT Transport Solutions

For accurate assessment of nuclear heating in fast reactors, a new coupled neutron and gamma heating calculation scheme has been developed based on VARIANT nodal transport solutions of neutron and gamma flux distributions. The MC2-3 code was extended to generate multigroup neutron and gamma cross sections and kinetic energy release in materials (KERMA) factors, and a utility program CURVE was developed to reconstruct detailed pin and duct wall powers from VARIANT output files. The improved heating calculation scheme has been verified against MCNP6 Monte Carlo reference solutions for the Advanced Burner Test Reactor (ABTR) and Experimental Breeder Reactor II (EBR-II) benchmark problems. Compared to the existing coupled heating calculation method based on DIF3D diffusion theory solutions, the new heating calculation scheme utilizes more accurate gamma cross sections and KERMA factors, accounts for the transport effects, and eliminates the approximations in the existing pin power reconstruction scheme. As a result, it produces more accurate assembly and pin power distributions. For both the ABTR and EBR-II problems, the maximum assembly power error was ~1% in fuel assemblies and ~2% in instrumented structure assemblies, and the maximum error in pin segment powers in an axial node of fuel assembly was ~4%. In the blankets of the EBR-II problem, the maximum error in pin segment powers was increased to ~8%, mainly due to the lower power level and the relatively large error in the nodal power of the VARIANT solution.

MC2-3: Multigroup Cross Section Generation Code for Fast Reactor Analysis

This paper presents the methods and performance of the MC2-3 code, which is a multigroup cross-section generation code for fast reactor analysis, developed to improve the resonance self-shielding and spectrum calculation methods of MC2-2 and to simplify the current multistep schemes generating region-dependent broad-group cross sections. Using the basic neutron data from ENDF/B data files, MC2-3 solves the consistent P1 multigroup transport equation to determine the fundamental mode spectra for use in generating multigroup neutron cross sections. A homogeneous medium or a heterogeneous slab or cylindrical unit cell problem is solved in ultrafine (2082) or hyperfine (~400 000) group levels. In the resolved resonance range, pointwise cross sections are reconstructed with Doppler broadening at specified temperatures. The pointwise cross sections are directly used in the hyperfine group calculation, whereas for the ultrafine group calculation, self-shielded cross sections are prepared by numerical integration of the pointwise cross sections based upon the narrow resonance approximation. For both the hyperfine and ultrafine group calculations, unresolved resonances are self-shielded using the analytic resonance integral method. The ultrafine group calculation can also be performed for a two-dimensional whole-core problem to generate region-dependent broad-group cross sections. Verification tests have been performed using the benchmark problems for various fast critical experiments including Los Alamos National Laboratory critical assemblies; Zero-Power Reactor, Zero-Power Physics Reactor, and Bundesamt für Strahlenschutz experiments; Monju start-up core; and Advanced Burner Test Reactor. Verification and validation results with ENDF/B-VII.0 data indicated that eigenvalues from MC2-3/DIF3D agreed well with Monte Carlo N-Particle5 MCNP5 or VIM Monte Carlo solutions within 200 pcm and regionwise one-group fluxes were in good agreement with Monte Carlo solutions.

A neutronic benchmark specification and COMET solution for the Advanced Burner Test Reactor

This work describes a stylized three dimensional benchmark problem based on Argonne National Laboratory’s conceptual Advanced Burner Test Reactor design. This reactor is a sodium cooled fast reactor designed to burn recycled fuel to generate power while transmuting transuranic waste. The specification includes heterogeneity at both the assembly and core levels while the geometry and material compositions are both fully described. Material nuclear data is provided in the form of a set of 15 group macroscopic cross sections which were developed so that the benchmark specification could be used for transport code method verification.Using this benchmark and 15 group cross sections, the Coarse Mesh Transport Method (COMET) code was verified against the Monte Carlo code MCNP. Results were generated for three separate core cases: control rods out, near critical, and control rods in. Eigenvalue comparison results between COMET and MCNP are within 50pcm for all three cases, well within two standard deviations of the stochastic uncertainty of the solution methods. The fission density results are highly accurate with a pin fission density average discrepancy between COMET and MCNP of less than 0.5% for each model.

Experiment Optimization to Reduce Nuclear Data Uncertainties in Support of Reactor Design

The safe and economical design of new, innovative nuclear reactors will require uncertainty reduction in basic nuclear data that are input to simulations used during reactor design. These data uncertainties propagate to uncertainties in design responses, which in turn require the reactor designer to incorporate additional safety margins into the design, often increasing the cost of the reactor. Therefore, basic nuclear data need to be improved, and this is accomplished through experimentation, which is often done using cold critical experiments. Considering the high cost of nuclear experiments, it is desired to have an optimized experiment that will provide the experimental data needed for maximum uncertainty reduction in the design responses. However, the optimization of the experiment is coupled to the reactor design itself because with reduced uncertainty in the design responses the reactor design can be re-optimized. It is thus desired to find the experiment design that gives the most optimized reactor design. Solution of this nested optimization problem is made possible by the use of the simulated annealing algorithm. Cost values for experiment design specifications and reactor design specifications are estimated and used to compute a total savings by comparing the a posteriori reactor cost to the a priori cost accounting for the offsetting cost of the experiment. This was done for the Argonne National Laboratory-developed Advanced Burner Test Reactor design concept employing a modified Zero Power Physics Reactor as the experimental facility.

Elevated temperature design evaluations for the ABTR internal structure

The preliminary structural analyses for an advanced burner test reactor (ABTR) internal structure are performed with regard to the elevated temperature design and the seismic design. In the elevated temperature design, three load cases enveloped from ABTR duty cycle events are studied for the structural integrity evaluation according to the ASME-NH design rule. The resulting stresses and strains are compared with the allowable design and service limits. An elastic approach, a simplified inelastic approach and a creep fatigue analysis are performed to provide a quantitative assessment of the deformations and strains. In the seismic design, the response spectrum analyses including the fluid-structure interaction effect are performed for isolated and non-isolated conditions. From the evaluation results, it is found that the seismic isolation capable of reducing the seismic input transmitted to the foundations is needed to satisfy the allowable design limits. For the isolated condition, three design cases for the variation of the wall thickness as 1.5 cm, 2 cm and 3 cm are studied for the design modification.

Buckling limit evaluations for reactor vessel of ABTR

We evaluated the buckling limit of a conceptually designed reactor vessel of the Advanced Burner Test Reactor (ABTR) for a horizontal safe shutdown earthquake (SSE) seismic load. In this evaluation, both seismic isolation and non-isolation designs were considered for thin reactor vessels subjected to elevated temperature services. For calculating the buckling load, two kinds of methods, a numerical simulation method using finite element analysis and an evaluation formula driven from a theoretical basis, were used. To consider the material aging effect caused by a 60-year design lifetime and 510°C normal operating temperature, an isochronous stress-strain curve corresponding to these conditions was used for the nonlinear elastic-plastic buckling analysis method. From the evaluation results of the buckling load, it was found that plasticity behavior significantly affects the buckling strength, but that the initial geometrical imperfections have little effect. Also, the non-seismic isolation design does not satisfy the buckling limit rules of the ASME BPV Section III, Subsection NH, but the seismic isolation design does satisfy it with sufficient margins.

Benchmark of Advanced Burner Test Reactor model using MCNPX 2.6.0 and ERANOS 2.1

Significant research is currently being performed whereby fast reactor cores have been designed to burn transuranic materials reducing the volume and long-term radiotoxicity of spent nuclear fuel. These core and depletion models depend on various computer codes. This research used MCNPX 2.6.0 and ERANOS 2.1 to model a standard 250 MW Advanced Burner Test Reactor (ABTR) core. The intent was to benchmark criticality and burnup results from a stochastic Monte Carlo code and a deterministic depletion code using a standard ABTR model created by Argonne National Laboratory. Because each of these codes solves the transport and burnup problem differently, there is a need to benchmark the core models in order to verify results and identify root causes for significant differences in results between codes. Flux calculations in ERANOS were performed using diffusion theory, Legendre polynomial approximations (using the VARIANT module) and discrete ordinates methods. The k-effective for the higher order transport models remained within 1000 pcm of the MCNPX model. The difference between the total heavy nuclide mass balance in ERANOS using the various flux calculations and the MCNPX depletion model was less than 0.4% out to a burnup of 1095 days (67.45 GWd/MTHM). The percent delta between the codes as a fraction of the fissioned mass was 1.34%. For the isotopes with large concentrations, such as 238U and 239Pu, the mass differences were 0.38% and 0.01% respectively. The mass difference for 241Am was also small at 0.42%. Notable isotopes in small quantities with larger mass differences were 242Am, 242Cm, 243Cm and 246Cm where differences ranged from 0.1 to 0.2% after 26 days and increased to 11–136% at 1095 days.

A CFD simulation process for fast reactor fuel assemblies

A CFD modeling and simulation process for large-scale problems using an arbitrary fast reactor fuel assembly design was evaluated. Three-dimensional flow distributions of sodium for several fast reactor fuel assembly pin spacing configurations were simulated on high performance computers using commercial CFD software. This research focused on 19-pin fuel assembly “benchmark” geometry, similar in design to the Advanced Burner Test Reactor, where each pin is separated by helical wire-wrap spacers. Several two-equation turbulence models including the k– ɛ and SST (Menter) k– ω were evaluated. Considerable effort was taken to resolve the momentum boundary layer, so as to eliminate the need for wall functions and reduce computational uncertainty. High performance computers were required to generate the hybrid meshes needed to predict secondary flows created by the wire-wrap spacers; computational meshes ranging from 65 to 85 million elements were common. A general validation methodology was followed, including mesh refinement and comparison of numerical results with empirical correlations. Predictions for velocity, temperature, and pressure distribution are shown. The uncertainty of numerical models, importance of high fidelity experimental data, and the challenges associated with simulating and validating large production-type problems are presented.

Scalable parallel solution coupling for multiphysics reactor simulation

Reactor simulation depends on the coupled solution of various physics types, including neutronics, thermal/hydraulics, and structural mechanics. This paper describes the formulation and implementation of a parallel solution coupling capability being developed for reactor simulation. The coupling process consists of mesh and coupler initialization, point location, field interpolation, and field normalization. We report here our test of this capability on an example problem, namely, a reflector assembly from an advanced burner test reactor. Performance of this coupler in parallel is reasonable for the chosen problem size and range of processor counts. The runtime is dominated by startup costs, which amortize over the entire coupled simulation. Future efforts will include adding more sophisticated interpolation and normalization methods, to accommodate different numerical solvers used in various physics modules and to obtain better conservation properties for certain field types.

Creep-fatigue design evaluations including daily load following operations for the advanced burner test reactor

In this paper, creep-fatigue design evaluations are carried out with representative duty cycle events for the advanced burner test reactor (ABTR), the sodium-cooled fast reactor (SFR), and the advanced recycling reactor (ARR) pre-conceptual design. Especially, the creep-fatigue damage by the duty cycle daily load follow operations are investigated with the detailed evaluations by the ASME-NH procedures. The main targeted component is the reactor vessel which has hot and cold pool free surface regions inducing significant thermal gradients at elevated temperatures. To reserve the creep-fatigue design margins, appropriate design modifications are investigated and proposed based upon the results of sensitivity studies. For the creep-fatigue evaluations, the SIE ASME-NH computer program, which implements the ASME-NH rules, is used with the results of the ANSYS elastic finite element analysis. From the results of the evaluations, the proposed modified design conditions satisfy the ASME-NH creep-fatigue limits with sufficient design margins. Especially, it was found that the daily load follow operations have an impact on the fatigue damages when compared with the other duty cycle events that were investigated but that the calculated values are so small as to be negligible, and the creep damages are slightly enhanced but they are also negligible.

Investigation of alternative layouts for the supercritical carbon dioxide Brayton cycle for a sodium-cooled fast reactor

Analyses of supercritical carbon dioxide (S-CO 2) Brayton cycle performance have largely settled on the recompression supercritical cycle (or Feher cycle) incorporating a flow split between the main compressor downstream of heat rejection, a recompressing compressor providing direct compression without heat rejection, and high and low temperature recuperators to raise the effectiveness of recuperation and the cycle efficiency. Alternative cycle layouts have been previously examined by Angelino (Politecnico, Milan), by MIT (Dostal, Hejzlar, and Driscoll), and possibly others but not for sodium-cooled fast reactors (SFRs) operating at relatively low core outlet temperature. Thus, the present authors could not be sure that the recompression cycle is an optimal arrangement for application to the SFR. To ensure that an advantageous alternative layout has not been overlooked, several alternative cycle layouts have been investigated for a S-CO 2 Brayton cycle coupled to the Advanced Burner Test Reactor (ABTR) SFR preconceptual design having a 510 °C core outlet temperature and a 470 °C turbine inlet temperature to determine if they provide any benefit in cycle performance (e.g., enhanced cycle efficiency). No such benefits were identified, consistent with the previous examinations, such that attention was devoted to optimizing the recompression supercritical cycle. The effects of optimizing the cycle minimum temperature and pressure are investigated including minimum temperatures and/or pressures below the critical values. It is found that improvements in the cycle efficiency of 1% or greater relative to previous analyses which arbitrarily fixed the minimum temperature and pressure can be realized through an optimal choice of the combination of the minimum cycle temperature and pressure (e.g., for a fixed minimum temperature there is an optimal minimum pressure). However, this leads to a requirement for a larger cooler for heat rejection which may impact the tradeoff between efficiency and capital cost. In addition, for minimum temperatures below the critical temperature, a lower heat sink temperature is required the availability of which is dependent upon the climate at the specific plant site.