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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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48 records · Page 3
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

SCALE Activities in FY23 [Slides]

This presentation shows that the SCALE 6.3 is available from RSICC. The lecture touches on new features including the new ENDF/B-VIII.0 data including covariances. It shows that the updated parallel infrastructure enables parallel capability on Windows. Production release with maintenance until 2026 at minimum addressing Code or data bugs, Performance, Ease of installation. There is a New Government Use Agreement (GUA) for SCALE 7.0 beta access. With site licenses available for non-commercial testing and feedback, handled through ORNL technology transfer.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2024

IER-499 Final Design of a TEX Variant: Chlorine to Support NNSA Operations [Slides]

This presentation concludes the final design of TEX with chlorine absorbers is complete. 5 configurations were identified, expected to down select to three experimental configurations. Great comparison to Y-12 and good comparison to INL sensitivities. Fastest configuration touches on the 35 Cl(n,p) cross section (continuing to work with Terrapower for faster configurations). Extensive density studies performed to ensure understanding of the salt in the plates. Reactivity studies performed to assist with experiment. Finally, the fabrication of the chlorine absorbers is ongoing (with LANL), and the experiment is scheduled to be completed this year.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

AMPX and SCALE Nuclear Data Libraries for Depletion [Slides]

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).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Zirconium Scattering Sensitivity in Neutron Transport Calculations of Multiplying Systems [Slides]

This presentation touches on motivation and applications of Zirconium. Performance of evaluated libraries on double and quasi differential experiments is considered. Additionally, the performance of evaluated libraries in select ICSBEP benchmarks is considered. Provided is a detailed look in the performance of evaluated libraries in preliminary benchmark configurations for IER-516: Zirconium Test Assembly (ZTA). Conclusions and recommended experiments to satisfy Zirconium scattering nuclear data needs are presented.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Can we safely go to 200 °C? An integrated approach to assessing impacts to the engineered barrier system in a high-temperature repository

Abstract. This presentation gives on overview of the complex thermo-hydro-mechanical-chemical (THMC) processes occurring during the disposal of heat-producing high-level radioactive waste in geologic repositories. A specific focus is on the role of compacted bentonite, which is commonly used as an engineered backfill material for emplacement tunnels because of its low permeability, high swelling pressure, and radionuclide retention capacity. Laboratory and field tests integrated with THMC modeling have provided an effective way to deepen our understanding of temperature-related perturbations in the engineered barrier system; however, most of this work has been conducted for maximum temperatures around 100 ∘C. In contrast, some international disposal programs have recently started investigations to understand whether local temperatures in the bentonite of up to 200 ∘C could be tolerated with no significant changes to safety relevant properties. Raising the maximum temperature is attractive for economical and safety reasons but faces the challenge of exposing the bentonite to significant temperature increases. Strong thermal gradients may induce complex moisture transport processes while geochemical processes, such as cementation and perhaps also illitization effects may occur, all of which could strongly affect the bentonite and near-field rock properties. Here, we present initial investigations of repository behavior exposed to strongly elevated temperatures. We will start discussing our current knowledge base for temperature effects in repositories exposed to a maximum temperature of 100 ∘C, based on data and related modeling analysis from a large heater experiment conducted for over 18 years in the Grimsel Test Site in Switzerland. We then show results from coupled THMC simulations of a nuclear waste repository in a clay formation exposed to a maximum temperature of 200 ∘C. We also explore preliminary data from a bench-scale laboratory mock-up experiment, which was designed to represent the strong THMC gradients occurring in a “hot” repository, and we finally touch on a full-scale field heater test to be conducted soon in the Grimsel Test Site underground research laboratory in Switzerland (referred to as HotBENT).

Birkholzer, Jens T. (ORCID:0000000279891912)↗
DOE OSTI2023

MCNP6 Developments: A 2022-23 Year in Review [Slides]

A separate MCNP6.3 V&V document reports on all the default calculations for all test suites is touched on. This report does not include the subcritical multiplication benchmark suite. After some additional clean-up and finalizing the post-processing and documentation steps, the subcritical multiplication benchmark suite will be released in the next version of our vnvstats repository. We tested the new HDF5 PTRAC feature in MCNP6.3 and found encouraging outcomes. Identical results coming out of the simulation with respect to the legacy PTRAC results. The overall runtime for all simulations is reduced by ~20% with the new HDF5 PTRAC capability. We consider giving the new HDF5 PTRAC features a try and using it for all subcritical multiplication and any other relevant (PTRAC) calculations.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2025

Nuclear data sensitivity and uncertainty study of copper-reflected integral experiments [Slides]

This presentation touches on reducing uncertainties in intermediate-energy actinide nuclear data and this continues to be a high priority for many applications. The goal of PARADIGM (PARallel Approach of Differential and InteGral Measurements): accelerate efforts to reduce biases and uncertainties in nuclear data through improvements to the nuclear data pipeline. This presentation includes integral experiments and a summarization of existing copper nuclear data.

Cu63↗
DOE OSTI2025

Code Coverage Status of the ARC Code PERSENT

The Argonne Reactor Code (ARC) software system supports users in their fast reactor design goals by providing neutronic, thermal-hydraulic, and structural analysis capabilities. PERSENT fulfills the role of generating reactivity coefficients for a given time point of a REBUS calculation usable in a point kinetics based safety analysis capability. PERSENT also provides a sensitivity coefficient capability on eigenvalue, reactivity worth, and several other key coefficients that are used in the follow-on safety analysis. Given a co-variance matrix, PERSENT can carry out the uncertainty quantification to indicate the amount of error in the reactivity coefficients derived from the errors in the cross section measurements. With continued improvement of computational resources, many of the geometry modeling capabilities in DIF3D that were primarily used in low order schemes are not really needed anymore. Today, the diffusion and transport capabilities of DIF3D-VARIANT are primarily used in the reactor design process with some scattered usage of DIF3D-FD and DIF3D-Nodal. PERSENT is part of the ARC code system and is built around DIF3D-VARIANT and the flux solution it provides. The purpose of the present work is to identify a set of test problems for PERSENT and assess the code coverage of PERSENT for those test problems. PERSENT treats the DIF3D executable as an external executable and thus the code coverage considerations only need to focus on the PERSENT source code and only a fraction of the connected modules in the existing ARC software library. The goal is to document what parts of the existing PERSENT code are touched by the set of test problems and which are not. Because the verification work done on PERSENT was focused on the most common uses of PERSENT for fast reactor analysis, the code coverage assessment of those capabilities is the highest priority. This will ensure that nothing is being missed by the existing verification test problems that users of PERSENT rely upon. The code coverage analysis of PERSENT was performed with the Code Coverage Tool of the Intel Fortran compiler which requires modifications to the compilation of PERSENT. The detailed coverage tables are given for each submodule of PERSENT. Most of the uncovered parts/files could be easily ignored because they are either for error message and debugging output or not needed by PERSENT today. Only a few uncovered parts of PERSENT deserve extending the verification test suite.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Fast Neutron Evaluations for Advanced Lead Systems [Slides]

This presentation touches on fast neutron evaluations for advanced lead systems. This lecture covers project goals to evaluate lead (Pb) isotopes with a focus on improving predictive capabilities of fast systems. Additionally, the project sets out to benchmark available critical and shielding experiments with sensitivity to Pb and utilize RPI Quasi-Differential Scattering experiment for scattering reactions. This project will perform Resolved Resonance Region (RRR) and fast region evaluations for natural Pb isotopes with differential data on EXchange FORmat for experimental numerical nuclear reaction data (EXFOR) and ultimately will submit evaluations to the ENDF/B-VIII.1 library.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

On Integral Benchmarks for Resolving TSL Conflicts for ENDF/B-VIII.1 Release [Slides]

This presentation touches on polystyrene ORNL evaluation in TSL review process for ENDF/B-VIII.1. the presentation found excellent agreement with INS and transmission measurements. Additionally, PS and PE evaluations with different phonon spectra, similar total xs., calculate keff similarly. This presentation found for critical benchmarks. While extremely useful for validation of nuclear data at all energies, critical benchmarks are not the best tool to provide a definitive answer on conflicting TSLs. In conclusion this presentation found INS and transmission measurements need to be the basis of validation of TSLs.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Dynamic Human-in-the-Loop Simulated Nuclear Power Plant Thermal Dispatch System Demonstration and Evaluation Study

An Idaho National Laboratory research team performed a human-in-the-loop study with two formerly licensed retired operators to evaluate thermal power dispatch operations supported by the modified GSE Systems GPWR plant simulator. Virtual representations of the analog control panels were presented on touch-screen bays configured to mimic the control room layout in the newly renovated Human Systems Simulation Laboratory. The operators performed 15 scenarios covering normal evolutions to transition the plant from full turbine operation to joint turbine and thermal power dispatch operations in addition to transient response scenarios induced with simulated faults to evaluate the impact of thermal power dispatch system on operator and plant responses. A prototype human-system interface (HSI) was developed and displayed in tandem with the virtual analog panels to support the operators executing the procedurally drive evolutions and transient responses. An interdisciplinary team of operations experts, nuclear engineers, and human factors experts observed the operators performing the scenarios to evaluate the operations. Only a preliminary analysis of the results has been performed. The full analysis will be shared in a milestone report scheduled for release at the end of September; however, some high-level conclusions were readily available. Two high level findings for the system design were captured in the study. The manual control supported by the HSI to transition from standard operations to thermal power dispatch operation imposed a considerable amount of workload on the operators due to tedious manual valve manipulations and system monitoring required to verify their intended effect. An additional operator would be required in the control room to support the daily evolution. Automatic control for the transition was deemed a requirement for plant adoption without imposing additional staffing costs. The second finding was the necessity for an automatic thermal power dispatch system trip isolation function linked to a turbine and reactor trip signal. The operators completed scenarios with automatic isolation functionality and manually required actuation of the thermal power dispatch system. The operator response was sufficiently slower in the manual trip condition, such that operators were unable to manually actuate key post trip safety functions an indicate a degraded control capability that should be avoided. With an automatic trip signal little to no impact of the thermal power dispatch system was identified on the primary plant response and therefore the system could be readily and safely adopted. Together these two findings represent the need to support the adoption of thermal power dispatch capability into existing operations by leveraging automation to augment any additional operator tasking required to control and monitor an additional system beyond existing operations.

99 GENERAL AND MISCELLANEOUS↗