Heuristic optimization of group structure using Physics-Based Fitness Approximation
Not provided.
Nuclear Science & Technology↗
Engineering topics
Sjoden, G.
Publications and source records attributed to Sjoden, G..
Development and application of a hybrid transport methodology for active interrogation systems
A hybrid Monte Carlo and deterministic methodology has been developed for application to active interrogation systems. The methodology consists of four steps: i) neutron flux distribution due to neutron source transport and subcritical multiplication; ii) generation of gamma source distribution from (n, 7) interactions; iii) determination of gamma current at a detector window; iv) detection of gammas by the detector. This paper discusses the theory and results of the first three steps for the case of a cargo container with a sphere of HEU in third-density water cargo. To complete the first step, a response-function formulation has been developed to calculate the subcritical multiplication and neutron flux distribution. Response coefficients are pre-calculated using the MCNP5 Monte Carlo code. The second step uses the calculated neutron flux distribution and Bugle-96 (n, 7) cross sections to find the resulting gamma source distribution. In the third step the gamma source distribution is coupled with a pre-calculated adjoint function to determine the gamma current at a detector window. The AIMS (Active Interrogation for Monitoring Special-Nuclear-Materials) software has been written to output the gamma current for a source-detector assembly scanning across a cargo container using the pre-calculated values and taking significantly less time than a reference MCNP5 calculation. (authors)
97 MATHEMATICS AND COMPUTING↗
Advanced quadratures and periodic boundary conditions in parallel 3D S{sub n} transport
Significant updates in numerical quadratures have warranted investigation with 3D Sn discrete ordinates transport. We show new applications of quadrature departing from level symmetric ( S{sub 2}o). investigating 3 recently developed quadratures: Even-Odd (EO), Linear-Discontinuous Finite Element - Surface Area (LDFE-SA), and the non-symmetric Icosahedral Quadrature (IC). We discuss implementation changes to 3D Sn codes (applied to Hybrid MOC-Sn TITAN and 3D parallel PENTRAN) that can be performed to accommodate Icosahedral Quadrature, as this quadrature is not 90-degree rotation invariant. In particular, as demonstrated using PENTRAN, the properties of Icosahedral Quadrature are suitable for trivial application using periodic BCs versus that of reflective BCs. In addition to implementing periodic BCs for 3D Sn PENTRAN, we implemented a technique termed 'angular re-sweep' which properly conditions periodic BCs for outer eigenvalue iterative loop convergence. As demonstrated by two simple transport problems (3-group fixed source and 3-group reflected/periodic eigenvalue pin cell), we remark that all of the quadratures we investigated are generally superior to level symmetric quadrature, with Icosahedral Quadrature performing the most efficiently for problems tested. (authors)
97 MATHEMATICS AND COMPUTING↗
SNM neutron detection using a time-gated synthetic aperture hybrid approach
This work focuses on using forward and adjoint transport in a hybrid application of 3-D deterministic (PENTRAN) and Monte Carlo (MCNP5) codes to model a series of neutron detector blocks. These blocks, or 'channels, ' contain a unique set of moderators with 4 atm He-3 detectors tuned to detect and profile a gross energy spectrum of a passing neutron (SNM) source. Ganging the units together as a large area system enables one to apply time gating the source-detector response to maximize signal to noise responses from a passing source with minimal background; multiple units may be positioned as a collective synthetic aperture detector array to be used as a way of performing real time neutron spectroscopy for detecting special nuclear materials in moving vehicles. The initial design, detector response coupling, confirmation of initial design functionality using adjoint transport calculations, and realistic simulation using PENTRAN and MCNP5 are presented. Future work will include optimization and application to realistic scenarios and additional sources. (authors)
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Optimization of a neutron detector design using adjoint transport simulation
A synthetic aperture approach has been developed and investigated for Special Nuclear Materials (SNM) detection in vehicles passing a checkpoint at highway speeds. SNM is postulated to be stored in a moving vehicle and detector assemblies are placed on the road-side or in chambers embedded below the road surface. Neutron and gamma spectral awareness is important for the detector assembly design besides high efficiencies, so that different SNMs can be detected and identified with various possible shielding settings. The detector assembly design is composed of a CsI gamma-ray detector block and five neutron detector blocks, with peak efficiencies targeting different energy ranges determined by adjoint simulations. In this study, formulations are derived using adjoint transport simulations to estimate detector efficiencies. The formulations is applied to investigate several neutron detector designs for Block IV, which has its peak efficiency in the thermal range, and Block V, designed to maximize the total neutron counts over the entire energy spectrum. Other Blocks detect different neutron energies. All five neutron detector blocks and the gamma-ray block are assembled in both MCNP and deterministic simulation models, with detector responses calculated to validate the fully assembled design using a 30-group library. The simulation results show that the 30-group library, collapsed from an 80-group library using an adjoint-weighting approach with the YGROUP code, significantly reduced the computational cost while maintaining accuracy. (authors)
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Hybrid parallel code acceleration methods in full-core reactor physics calculations
When dealing with nuclear reactor calculation schemes, the need for three dimensional (3D) transport-based reference solutions is essential for both validation and optimization purposes. Considering a benchmark problem, this work investigates the potential of discrete ordinates (Sn) transport methods applied to 3D pressurized water reactor (PWR) full-core calculations. First, the benchmark problem is described. It involves a pin-by-pin description of a 3D PWR first core, and uses a 8-group cross-section library prepared with the DRAGON cell code. Then, a convergence analysis is performed using the PENTRAN parallel Sn Cartesian code. It discusses the spatial refinement and the associated angular quadrature required to properly describe the problem physics. It also shows that initializing the Sn solution with the EDF SPN solver COCAGNE reduces the number of iterations required to converge by nearly a factor of 6. Using a best estimate model, PENTRAN results are then compared to multigroup Monte Carlo results obtained with the MCNP5 code. Good consistency is observed between the two methods (Sn and Monte Carlo), with discrepancies that are less than 25 pcm for the k{sub eff}, and less than 2.1% and 1.6% for the flux at the pin-cell level and for the pin-power distribution, respectively. (authors)
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Source term characterization for SNM pit storage facilities
In order to properly design a mobile system to validate and verify the presence of special nuclear materials for non-proliferation and safeguards applications, accurate modeling of source materials is imperative. In this work, models were developed for use in design assessments based on an AL-R8 SNM standardized container specification to determine the radioactive signatures for both highly enriched uranium (HEU) and weapons plutonium (WGPu) special nuclear materials (SNM) housed in the containers. Intrinsic gamma boundary leakage currents were evaluated for this system, performed using 3D fixed-source deterministic SN photon transport (PENTRAN) as well as with stochastic Monte Carlo methods (MCNP5). Group-dependent leakage radiation terms were calculated at two 'source box' interfaces within the models, one directly surrounding the SNM source, and one immediately surrounding the canister. Analysis showed good agreement between the two models for energy groups of interest, based on a 24 group gamma library established for HEU and WGPu gamma signatures of interest. Intrinsic and neutron induced gamma leakage was determined using Monte Carlo calculations, and the combined gamma signatures were then treated as a net gamma leakage to be used in subsequent photon transport calculations. Neutron leakage based on the BUGLE-96 47 group structure was determined using Monte Carlo calculations for the WGPu canisters. These results will be used to evaluate the source term from stored nuclear materials and augment our efforts to design a detection system to validate the presence of these materials for safeguards purposes. (authors)
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗