Engineering topics

Celik, Cihangir

Publications and source records attributed to Celik, Cihangir.

58 records · Page 4

DOE OSTI2013

Overview of Scale 6.2

SCALE is an industry-leading suite of tools for nuclear systems modeling and simulation that provides comprehensive, verified and validated, user-friendly capabilities for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. For more than 30 years, regulators, licensees, and research institutions around the world have used SCALE for nuclear safety analysis and design. SCALE provides a plug-and-play framework that includes three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution. SCALE includes the latest nuclear data libraries for continuous-energy and multigroup radiation transport as well as activation, depletion, and decay calculations. SCALE s graphical user interfaces assist with accurate system modeling, visualization, and convenient access to desired results. SCALE 6.2 provides several new capabilities and significant improvements in many existing features, especially with expanded CE Monte Carlo capabilities for criticality safety, shielding, depletion, and sensitivity and uncertainty analysis. A brief overview of SCALE capabilities is provided with emphasis on new features for SCALE 6.2.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2013

Enhancements in Continuous-Energy Monte Carlo Capabilities in SCALE

Monte Carlo tools in SCALE are commonly used in criticality safety calculations as well as sensitivity and uncertainty analysis, depletion, and criticality alarm system analyses. Recent improvements in the continuous-energy data generated by the AMPX code system and significant advancements in the continuous-energy treatment in the KENO Monte Carlo eigenvalue codes facilitate the use of SCALE Monte Carlo codes to model geometrically complex systems with enhanced solution fidelity. The addition of continuous-energy treatment to the SCALE Monaco code, which can be used with automatic variance reduction in the hybrid MAVRIC sequence, provides significant enhancements, especially for criticality alarm system modeling. This paper describes some of the advancements in continuous-energy Monte Carlo codes within the SCALE code system.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2013

Hybrid Technique in SCALE for Fission Source Convergence Applied to Used Nuclear Fuel Analysis

The new hybrid SOURCE ConveRgence accelERator (SOURCERER) sequence in SCALE deterministically computes a fission distribution and uses it as the starting source in a Monte Carlo eigenvalue criticality calculation. In addition to taking the guesswork out of defining an appropriate, problem-dependent starting source, the more accurate starting source provided by the deterministic calculation decreases the probability of producing inaccurate tally estimates associated with undersampling problems caused by inadequate source convergence. Furthermore, SOURCERER can increase the efficiency of the overall simulation by decreasing the number of cycles that has to be skipped before the keff accumulation. SOURCERER was applied to a representative example for a used nuclear fuel cask utilized at the Maine Yankee storage site {Scaglione and Ilas}. Because of the time constraints of the Used Fuel Research, Development, and Demonstration project, it was found that using more than 30,000 neutrons per cycle will lead to inaccurate Monte Carlo calculation of keff due to the inevitable decrease in the number of skipped and active cycles used with this problem. For a fixed uncertainty objective and by using 30,000 neutron per cycle, the use of SOURCERER increased the efficiency of the keff calculation by 60%compared to a Monte Carlo calculation that used a starting source distributed uniformly in fissionable regions, even with the inclusion of the extra computational time required by the deterministic calculation. Additionally, the use of SOURCERER increased the reliability of keff calculation using any number of skipped cycles below 350.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2013

Validation of Criticality Safety Calculations with SCALE 6.2

SCALE 6.2 provides numerous updates in nuclear data, nuclear data processing, and computational tools utilized in the criticality safety calculational sequences relative to SCALE 6.1. A new 252-group ENDF/B-VII.0 multigroup neutron library, improved ENDF/B-VII.0 continuous energy data, as well as the previously deployed 238-group ENDF/B-VII.0 neutron library are included in SCALE 6.2 for criticality safety analysis. The performance of all three libraries for keff calculations is examined with a broad sampling of critical experiment models covering a range of fuels and moderators. Critical experiments from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (IHECSBE) that are available in the SCALE Verified, Archived Library of Inputs and Data (VALID) are used in this validation effort. Over 300 cases are used in the validation of KENO V.a, and a more limited set of approximately 50 configurations are used for KENO-VI validation. Additionally, some KENO V.a cases are converted to KENO-VI models so that an equivalent set of experiments can be used to validate both codes. For continuous-energy calculations, SCALE 6.2 provides improved performance relative to SCALE 6.1 in most areas with notable improvements in fuel pin lattice cases, particularly those with mixed oxide fuel. Multigroup calculations with the 252-group library also demonstrate improved performance for fuel lattices, uranium (high and intermediate enrichment) and plutonium metal experiments, and plutonium solution systems. Overall, SCALE 6.2 provides equivalent or smaller biases than SCALE 6.1, and the two versions of KENO provide similar results on the same suite of problems.

97 MATHEMATICS AND COMPUTING↗