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This presentation covers the status of GNDS Support in AMPX. This presentation covers an overview of GNDS support in AMPX. Then the presentation touches on the AMPX process in GNDS and a checklist. This presentation concludes with a continuing effort section and conclusions.
This presentation provides an overview of ENDF reading in AMPX, as well as information regarding low level access classes for GNDS used in AMPX and the use of these classes in AMPX and SAMMY.
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.
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.
The modernization of SAMMY continued with the removal of all container array access for the main program. The container array was used due to the computer limitations at the time of the initial writing of SAMMY, but it prevented the use of some modern code testing options. Additionally, the authors continued to make the code more modular by shifting the storage of the energy grid and the covariance information to C++ storage, thus allowing for better unit testing and the elimination of scratch files. For AMPX, the authors continued to implement the new Evaluated Nuclear Data File format to allow AMPX to read new evaluations as they become available.
This presentation is on ENDF/B-VIII.1 and AMPX. This presentation begins with a brief introduction to the AMPX nuclear data processing code. The lecture then touches on nuclear data libraries that ship with the SCALE code system. Finally, this lecture concludes on preparations for the ENDF/B-VIII.1 libraries.
The 2020 edition of The International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook includes a newly produced plastic-moderated evaluation, identified as PU-MET-MIXED-002 and referred to as PMM002. The 2021 edition should include another new plastic-moderated experiment, HEU-MET-THERM-004, referred to as HMT-004. These evaluations are of particular interest, as PMM-002 is moderated with polyethylene, and HMT-004 is moderated with polymethyl methacrylate (Lucite), allowing for investigation of the differences in the thermal scattering laws (TSLs) available for these materials. ENDF/B-VIII.0 includes newly produced data for 1 H-based scattering materials, including Lucite and yttrium hydride. Other available 1 H TSLs include polyethylene, light water, zirconium hydride. The unbound 1 H (free-gas) cross sections were also used. ENDF/B-VIII.0 also includes revisions made to the ENDF/B-VII.1 light water and polyethylene evaluations. Zirconium hydride was unchanged from ENDF/B-VII.1. As of this writing, there are no benchmarks included in the Verified, Archived Library of Inputs and Data (VALID) that are primarily moderated with any solid moderator, so validation that includes the PMM-002 and HMT-004 evaluations would expand the coverage to two new moderators. A study was undertaken at Oak Ridge National Laboratory with two purposes: to provide validation data based on systems that are primarily moderated with polyethylene and Lucite, and to demonstrate the reactivity changes that can result from use of different 1 H TSLs in these plastic-moderated systems. The first objective was mainly to test data and code for SCALE and AMPX, whereas the second objective was to provide useful data for practitioners on the potential variability of predicted $k_{eff}$ based on TSL changes. The use of an exactly correct TSL is often not possible given the materials involved (e.g., lubricants), but this study was intended to provide some indication of the magnitude of the changes among similar materials that can manifest through application of different TSLs in hydrogenous systems over a range of different neutron energy spectra. Unfortunately, the nominal $k_{eff}$ results exposed deviations between SCALE and Monte Carlo N-Particle (MCNP). Some of the results using different TSLs were also unexpected and difficult to explain. An investigation revealed a processing issue in AMPX that was caused by an ambiguous description of the data for incoherent elastic scattering in the ENDF manual. This issue is discussed in detail in a SCALE User Notification, and it affects some solid moderators, but it does not apply to 1 H bound in water. This issue highlights the importance of validating all TSLs used in safety analysis calculations. The remainder of this paper provides a more detailed examination of the results that triggered the investigation and a summary of the findings of that investigation.
The modernization of SAMMY continued with consolidation of access to covariance information for adjusted parameters and data in SAMMY. This consolidation allowed for removal of many scratch files. In addition, work was initiated to make the 0K cross section calculation more modular and less dependent on SAMMY global parameters. An initial application programming interface (API) was added to expose cross sections (including resolution broadening) generated by SAMMY to external fitting routines. The processing for thermal moderators in AMPX was updated for selected moderators for which the generated grid was not fine enough. Updated libraries were generated for SCALE. In addition, work continued to fully support new Evaluated Nuclear Data File (ENDF) formats, including the Generalized Nuclear Database Structure (GNDS) in AMPX.
This presentation provides overviews of ENDF and GNDS reading in AMPX, AMPX covariance processing updates, thermal scattering law updates, and data libraries for SCALE 6.3.0.
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 presentation starts off with covering general ENDF issues encountered. With a look into issues particular to thermal scattering library. This presentation includes AMPX Development Highlights. The talk concludes with the status of the AMPX Photonuclear Sublibrary Support.
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.
The priority for AMPX development for FY2024 has been to support processing the ENDF/B-VIII.1 library. This includes the development to support Thermal Neutron Scattering Library, GNDS-2.1 support and AMPX maintenance, and transport libraries for SCALE 7.0.0.
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 study was conducted to validate the SCALE/Polaris v6.3.0–PARCS v3.4.2 code procedure with the Evaluated Nuclear Data File (ENDF)/B-VII.1 AMPX 56-group library for pressurized water reactor (PWR) analysis, by comparing simulated results with measured data for critical experiments and operating PWRs. Uncertainties of the SCALE/Polaris–PARCS code procedure for PWR analysis were evaluated in the validation for the PWR key nuclear parameters such as critical boron concentrations, reactivity, control bank work, temperature coefficients, and pin and assembly power peaking factors.
The SCALE/Polaris–PARCS code procedure has been used in the confirmatory analysis for boiling water reactors by the US Nuclear Regulatory Commission. In this study, the SCALE/Polaris v6.3.0–PARCS v3.4.2 code procedure with the Evaluated Nuclear Data File (ENDF)/B-VII.1 AMPX 56-group library was validated by comparing the simulated results with the measured data for operating boiling water reactors, including Peach Bottom Unit 2 cycles 1–3, Hatch Unit 1 cycles 1–3, and Quad Cities Unit 1 cycles 1–3. The uncertainties and biases of the SCALE/Polaris–PARCS code package for boiling water reactor physics analysis were evaluated in the validation for key nuclear parameters such as reactivity and traversing in-core probe data.
The AMPX nuclear data processing code reads evaluated nuclear data files and prepares libraries in continuous energy (CE) and multigroup (MG) formats. These libraries are used by the transport and depletion codes in the SCALE code system.