Engineering topics
Publications and source records attributed to Jeon, B. K..
This presentation focus on AMPX development for 2024 it also touches on ENDF/B-VIII.1. The slides touches on Thermal Neutron Scattering Library, AMPX Development Highlights, and Transport libraries for SCALE 7.0.0.
This presentation deals with AMPX development in FY2023 focused on preparing for the release and testing of ENDF/B-VIII.1. It includes updates for Thermal Scattering Law in ENDF/B-VIII.1. The presentation finishes with touching on resolution of a SCALE Issue with thermal kernels and photonuclear physics support.
This presentation provides an enlightening discussion of AMPX and ENDF/B-VIII.1. The talk begins with general ENDF issues encountered. Then narrows down into Issues particular to thermal scattering library. Such as the need for clear communication of the intended use for the file and the question of “Which graphite should I use in model?”. The presentation then looks at TSL files with many Bragg peaks are driving a change in SCALE CE library format and transport codes. This lecture concludes with a look at usage of thermal kernels with sharp peaks with the issues encountered using ENDF/B-VIII.0 data and resolved.
This presentation touches on AMPX and SCALE nuclear data libraries for depletion. AMPX is a nuclear data processing code suite that generates continuous-energy and multigroup neutron and gamma libraries used in SCALE and VERA, covariance data used in S/U calculations, and depletion, activation and decay origin data for Oak Ridge Isotope Generation code (ORIGEN).
This talk centers on AMPX developments in fiscal year of 2022. This presentation provides links to the open-source subset of SCALE, including AMPX. The discussion additionally provides an overview of GNDS support in AMPX. The presentation includes selected features on several SCALE versions and features.
This presentation covers recent AMPX developments. Presentation highlights include the news of AMPX being openly available, link to source and configure options are included. The interconnections between AMPX and Generalized Nuclear Database Structure (GNDS) are additionally highlighted as well.
This presentation discusses and announces that AMPX is now available as open source. It additionally provides an overview about ENDF and GNDS reading in AMPX. It also discusses thermal scattering law updates and the CE and MG data libraries for SCALE 6.3.0 and SCALE 7.0.0. In addition, the presentation discusses the photonuclear sub library processing capability of AMPX.
Polaris is a two-dimensional (2D) transport lattice physics capability in the SCALE code system for light-water reactor (LWR) analysis. SCALE/Polaris is being extended to treat doubly heterogeneous particulate fuels in LWRs with accident-tolerant fuels and prismatic high-temperature gas-cooled reactor (HTGR) fuel. Recently, a Dancoff-based Wigner-Seitz approximation capability based on the embedded self-shielding method (DWA-ESSM) was implemented into Polaris for efficient computation. A new double heterogeneity treatment capability based on DWA-ESSM in conjunction with the Hebert's collision probability method for double heterogeneity was developed and implemented for Polaris. The new capability was verified by performing benchmark calculations for the prismatic HTGR fuel compact problems with various design parameters through a code-to-code comparison between the Polaris and continuous-energy Monte Carlo results. (authors)
For the accurate assessment of the heat generation rate in fast reactors, the gamma library of MC2 -3 and the MC2 -3 + GAMSOR procedure employing the coupled neutron and gamma heating calculation has been thoroughly verified against Monte Carlo results and validated using the ZPPR-15D gamma dose measurement data. NJOY outputs are post-processed in a consistent way with the NJOY procedure to avoid any missing data or double counting of data. Prompt heating for 379 out of 391 isotopes in the gamma library was verified against MCNP6.2 to within 1% relative error in total heating for most isotopes. For both a simple one-dimensional slab problem representing a sodium cooled fast reactor and the Experimental Breeder Reactor II (EBR-II) benchmark problem, the root-mean-square values of assembly power error were less than 0.5% for fuel assemblies, ~1% in blankets and ~1 to ~3% in reflectors compared to MCNP6.2 results. The most plausible cause for the 3% error in a reflector assembly is believed to be the error in the multigroup neutron cross sections for the reflector assembly. For validation, gamma doses measured with thermoluminiscent dosimeters (TLDs) in the ZPPR-15D experiment were calculated using GAMSOR. Due to the uncertainty in the TLD measurement with regards to the energy deposition of photons and neutrons, the validation data leads to a 12.7% uncertainty on the experimental measurement. With this uncertainty bound, the calculated doses all fell within one standard deviation of the measured value. Combined with the accurate calculation of reaction rate distributions and neutron spectrum measurements, these results indicate good agreement for the neutron and gamma heating calculations that were performed.
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This document presents a detailed description of generating the gamma library for MC 2 -3 which includes delayed gammas and betas based on the 94 gamma energy groups. The updated methodologies associated with heating and gamma calculations are introduced, which account for the self-shielding effect on KERMA factors and perform the gamma transport calculations to condense the gamma data from 94 to 21 groups. For these updates, the PreGAMMA code was developed to prepare NJOY inputs and generate delayed gammas and betas, and the GenGAMMA code was updated to process the outputs of PreGAMMA and NJOY. This document briefly discusses verification results of the new gamma library using the ABTR fuel pin and the EBR-II benchmark problem.