SEARCH · Engineering Papers
Results for “mathematical physics”
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.
Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.
Polaris Depletion Calculations: Selected Best Practices [Slides]
Abstract not provided.
Sensitivity/Uncertainty Analysis for Nuclear Criticality Safety Validation [Slides]
Sensitivity coefficients predict the expected change in a response ($\kappa$) due to a change in some input parameter ($\Sigma$). These responses can be k eff (or ratios of reaction rates). These input parameters are typically nuclear data. The coefficients are dimensionless ratios. What would happen to the system k eff if some piece of data were changed by some amount? The coefficient is calculated without making the change.
Calculation of Spent Fuel Pool Time-to-Boil using Polaris/ORIGAMI [Slides]
The goal of this experiment was to calculate time-to-boil (TTB) for a Spent Fuel Pool (SFP). TTB is important for SFP management that represents the amount of time available to address loss of forced cooling. Researchers calculated and compared TTB for two cases using SCALE: SFP containing representative LEU fuel (≤ 5 wt %) and SFP containing representative LEU+ fuel (5 – 8 wt %). There were approximately 2300 fuel assemblies in the SFP for each case. Researchers concluded that the impact of LEU+ core on time-to-boil is very small. At such low cooling times, the impact of increased decay heat is not significant. Decay heat of SFP assemblies is small compared to discharge core decay heat.
Illustration of Neutronic Decoupling in Fresh Fuel Storage Environments [Slides]
Operations beyond static storage of fuel assemblies occur in SFPs (moving around the pool, inspection, etc.). These assemblies are assumed suitably distanced at 30 cm. When assemblies are separated by enough water, the neutrons from one cannot cause fission in another, which results in neutronic decoupling. This assumption was investigated for PWR and BWR assemblies as part of ORNL-TM/2021/2330.
Criticality and Radiation Shielding Impacts of ENDF/B-VIII.0 [Slides]
Several impacts of ENDF/B-VIII.0 on Radiation Shielding Applications were observed by researchers. Most reaction rates, attenuation ratios, and dose rates are similar for ENDF/B-VII.1 and ENDF/B-VIII.0. A significant reduction in reactor dosimetry results was noted. Investigation continues to reveal underlying data for discrepancies.
SCALE Sensitivity Tutorial [Slides]
In sensitivity analysis, we seek to quantify the degree to which fundamental data (i.e., nuclear data) influence a system’s response. In uncertainty analysis, we seek to quantify the degree to which uncertainty in fundamental data contributes to uncertainty in a system’s response. In SCALE, the TSUNAMI suite provides tools for sensitivity and uncertainty analysis, similarity analysis, and nuclear data and covariance adjustment.
Cumulative χ 2 Metric for VALID for ENDF/B-VII.1 and ENDF/B-VIII.0 in SCALE 6.3b9 [Slides]
Quantifying the performance of a new evaluation is difficult and doing so for a new library is significantly more challenging. Cumulative χ 2 metric is an established approach to comparing distributions. In this case, C/E results from different libraries. Andrei Trkov and Roberto Capote have published cumulative χ 2 results for ENDF/B-VIII.0. Cumulative χ 2 metric was recently calculated for VALID library maintained at ORNL.
Bayesian Monte-Carlo Evaluation Framework for Imperfect Nuclear Data [Slides]
BMC evaluation is a tool used to address imperfect data and models, non-linear models, and non-normal PDFs. ENDF-6 format does not allow non-normal parameter PDFs. Storing posterior sets allows for variance, covariance, skewness, etc. To better predict criticality, we should document non-normal parameter PDFs (i.e. asymmetric uncertainty) and consider non-linear sensitivity of $k_{\text{eff}}$ to resonance parameters.
Bayesian Monte-Carlo Framework: New Methods for Resonance Parameter Evaluation [Slides]
BMC evaluation is a tool to address imperfect data & models, non-linear models, and non-normal PDFs. ENDF-6 format does not allow non-normal parameter PDFs. Storing posterior sets allow for variance, covariance, skewness, etc. To better predict criticality, we could document non-normal parameter PDFs (i.e. asymmetric uncertainty), consider non-linear sensitivity of $\kappa$ eff to resonance parameters, and reduce uncertainty in key resonance parameters.
ORNL Neutron Cross Section Measurements of 90 Zr NCSP ND-1 Task [Slides]
Luiz Leal's idea to integrate the experimental uncertainties, covariances, and sensitivities in this workflow and centralize the process at one location led to the creation of a cohesive tool for data testing. It has been practiced for more than two decades by the NCS and ND groups at ORNL. The development and use of the ORNL codes TSUNAMI and TSURFER within SCALE made it possible. Today, it has been extended to SAMINT which enables the coupling of differential and integral data evaluation in a continuous-energy framework.
Advances in Nuclear Data Evaluation Theory for NCSP [Slides]
The generalized form of the Bayes’ Theorem improves UQ for NCSP of differential and/or integral data evaluations and of applications. It is being implemented in an API for use by SCALE, SAMMY, etc. Extended R-matrix formalism improves evaluations of RRR including direct reactions, direct capture, and the Reich-Moore approx., doorway state reactions with applications to RPI data on lead isotopes. Extended R-Matrix formalism will be implemented into SAMMY post modernization.
AMPX and SAMMY Status Report [Slides]
Abstract not provided.
Horizontal Split Table Conceptual Design for Validation of Nuclear Data used in Advanced Reactors [Slides]
A methodology to create conceptual designs of benchmark critical experiments for advanced reactors nuclear data testing and validation was developed. A first concept was explored, pebble-bed high-temperature gas cooled reactor, based on the HTR-10 reactor. The very high correlation is a proof of concept that the design is similar to the application, and performing such critical experiments would help nuclear data testing and validation. This proof-of-concept was included in the IER-539 CED-1:Preliminary Design of a New Horizontal Split Table report. Other concepts could be explored if needed, such as Molten-salt reactor, Sodium-cooled fast reactor or Heat pipe reactors/Microreactor.
Expanded Validation of Uranium Systems with the KENO Monte Carlo Codes and SCALE 6.2.4 [Slides]
There is likely a trend in bias for uranium-fueled systems as a function of spectrum, but probably not as a function of enrichment. Some enrichment trends appear, but are potentially the result of outlier experimental results; further investigation is needed where possible. There are good examples in this data of markedly different biases from different systems. Future work includes review and vetting for inclusion in the VALID library and analysis with TSUNAMI.
Challenges for Data Adjustment and Bias Quantification [Slides]
Abstract not provided.
SAMMY Modernization Efforts [Slides]
This presentation details several SAMMY modernization and maintenance goals, as well as the efforts made towards these goals and future plans for SAMMY.