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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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At least 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Uncertainty Analysis for a Criticality Benchmark [Slides]

This presentation discusses how KRUSTY was analyzed with LANL’s MCNP neutron transport code, Monte Carlo N-Particle. In the Monte Carlo method, neutrons are born in fission events and fly around the geometry having events (e.g., scatter, capture, fission, escape) according to the natural probabilities. “Natural probabilities” are given as neutron cross sections , which are measured and evaluated nuclear data. One of the goals of the benchmark project is to improve the nuclear cross sections (and the codes that use them). In this work, the response of interest is called k eff , which describes the neutron multiplication in a near critical system. Like most quantities, k eff is not measured directly; it is only inferred. In summation, MCNP results for k eff matched the measurements extremely well. The evaluated uncertainties are asymmetric and smaller than 0.00100 (e.g., + 0.00080 / – 0.00062). The largest uncertainties are due to alignment.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Port Gamble S'Klallam Tribe Strategic Energy Plan

The Port Gamble S'Klallam Tribe's Strategic Energy Plan (SEP) details the outcomes and efforts of a 1.5-year Energy Technology Innovation Partnership Program (ETIPP) project. This collaboration involved the Port Gamble S'Klallam Tribe, Spark Northwest, the Pacific Northwest National Laboratory (PNNL), and the National Renewable Energy Laboratory. Project partners supported the Tribe in identifying key energy research focus areas and the Tribe's energy vision. As the technical lead on the project, PNNL led a series of research initiatives to evaluate potential pathways for the Tribe to achieve its energy goals, subsequently providing associated recommendations. The SEP offers a comprehensive analysis encompassing several critical aspects including: -The Tribe's current energy landscape -Availability of renewable resources -Potential microgrid designs, developed by Säzän Group -Critical infrastructure analysis -Electric vehicle integration -An energy workforce development framework, led by Spark Northwest The analysis results come with recommended next steps, categorized by implementation timeframe (short-term, medium-term, long-term) and priority.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
NASA NTRS1980

Coal gasification systems engineering and analysis. Appendix F: Critical technology items/issues

Critical technology items and issues are defined in which there is a need for developmental research in order to assure technical and economic success for the state of the art of coal gasification in the United States. Technology development needs for the main processing units and the supporting units are discussed. While development needs are shown for a large number of systems, the most critical areas are associated with the gasifier itself and those systems which either feed the gasifier or directly receive products form the gasifier.

Source record↗
NASA NTRS2004

Gender and Decompression Sickness: A Critical Review and Analysis

The author addressed the following questions: are women at greater risk of decompression sickness and venous gas emboli at certain times in their reproductive cycle, is risk modified by the use of birth control pills (BCP), and is there a difference in overall risk between men and women under the same decompression dose? The summary considers information from the few abstracts and reports that were available. Except for the observation of more Type II DCS in women, particularly in women who fly after diving, there was no compelling evidence of a difference in DCS risk between men and women SCUBA divers. Many women that presented with DCS symptoms seemed to be in or near menses, with statistically fewer cases reported as time increased from menses. There was no compelling evidence that the use of BCP in SCUBA divers increases the risk of DCS. There were insufficient data about VGE from SCUBA diving to make any conclusion about the incidence of VGE and gender. In contrast, there were ample data about VGE from research in altitude chambers. Women produced less VGE and less Grade IV VGE compared to men under the same decompression dose, certainly when resting oxygen prebreathe (PB) was performed prior to ascent to altitude. Dual-cycle ergometry exercise during PB tends to reduce the differences in VGE between men and women. There was no compelling evidence that the risk of altitude DCS was different between men and women. However, a large number of DCS cases were associated with menses, and the use of BCP did seem to put women at a slightly greater risk than those that did not use BCP. There were substantial observations that women comprised a larger number of difficult cases that required complicated medical management.

Source record↗
NASA NTRS2017

Security Vulnerability Profiles of Mission Critical Software: Empirical Analysis of Security Related Bug Reports

While some prior research work exists on characteristics of software faults (i.e., bugs) and failures, very little work has been published on analysis of software applications vulnerabilities. This paper aims to contribute towards filling that gap by presenting an empirical investigation of application vulnerabilities. The results are based on data extracted from issue tracking systems of two NASA missions. These data were organized in three datasets: Ground mission IVV issues, Flight mission IVV issues, and Flight mission Developers issues. In each dataset, we identified security related software bugs and classified them in specific vulnerability classes. Then, we created the security vulnerability profiles, i.e., determined where and when the security vulnerabilities were introduced and what were the dominating vulnerabilities classes. Our main findings include: (1) In IVV issues datasets the majority of vulnerabilities were code related and were introduced in the Implementation phase. (2) For all datasets, around 90 of the vulnerabilities were located in two to four subsystems. (3) Out of 21 primary classes, five dominated: Exception Management, Memory Access, Other, Risky Values, and Unused Entities. Together, they contributed from 80 to 90 of vulnerabilities in each dataset.

Goseva-Popstojanova, Katerina↗
NASA NTRSDate not supplied

Critical Progressive Damage Analysis Model Features for Low-Velocity Impact on Thermoset and Thermoplastic Panels

A previous verification and validation framework evaluated the ability of various progressive damage analysis (PDA) tools to predict the mechanical response and damage state in post-buckled stiffened aerospace structures composed of fiber-reinforced thermoset composite materials. Herein, the ability of PDA tools to predict damage in thermoplastic composite materials subject to low-velocity impact (LVI) loading has been assessed. LVI tests and analyses have been conducted for similar panels, comprising thermoset or thermoplastic composite materials. Preliminary test/analysis correlations revealed that using the same combination of PDA model features and finite element model-building guidelines for the thermoset and thermoplastic LVI specimens resulted in underprediction of matrix crack lengths and delaminated area in the thermoplastic analyses despite yielding good correlations for the thermoset specimens. In this paper, the quality of the test/analysis correlations for analyses of thermoplastic LVI specimens performed with different combinations of PDA model features using the NASA CompDam material model are presented. Model features that were assessed in terms of effect on the quality of the test/analysis correlation include the representation of damageable interfaces, residual thermal stresses, and fiber damage.

Frank Leone↗
NASA NTRSDate not supplied

Critical Progressive Damage Analysis Model Features for Low-Velocity Impact on Thermoset and Thermoplastic Panels

A previous verification and validation framework evaluated the ability of various progressive damage analysis (PDA) tools to predict the mechanical response and damage state in post-buckled stiffened aerospace structures composed of fiber-reinforced thermoset composite materials. Herein, the ability of PDA tools to predict damage in thermoplastic composite materials subject to low-velocity impact (LVI) loading has been assessed. LVI tests and analyses have been conducted for similar panels, comprising thermoset or thermoplastic composite materials. Preliminary test/analysis correlations revealed that using the same combination of PDA model features and finite element model-building guidelines for the thermoset and thermoplastic LVI specimens resulted in underprediction of matrix crack lengths and delaminated area in the thermoplastic analyses despite yielding good correlations for the thermoset specimens. In this paper, the quality of the test/analysis correlations for analyses of thermoplastic LVI specimens performed with different combinations of PDA model features using the NASA CompDam material model are presented. Model features that were assessed in terms of effect on the quality of the test/analysis correlation include the representation of damageable interfaces, residual thermal stresses, and fiber damage.

Frank Leone↗
DOE OSTI2022

Analysis of SCALE Criticality and Sensitivity Calculations for Reflected HEU Cylinders [Abstract]

The SCALE code package offers several nuclear data libraries to support Monte Carlo (MC) transport, as well as MC-based derivation of $\kappa$ eff sensitivity and uncertainty (S/U) data. The CSAS sequence using the KENO MC code can utilize continuous-energy (CE) cross sections, or pre-generated multigroup (MG) cross section libraries. The use of MG libraries introduces bias into calculations in exchange for faster transport solutions. The TSUNAMI-3D sequence also utilizes KENO MC calculations. TSUNAMI-3D has two CE calculational methods: the Iterated Fission Probability (IFP) method, and the Contribution-Linked eigenvalue sensitivity/Uncertainty estimation via Tracklength importance CHaracterization (CLUTCH) method. Previous work has shown poor agreement between CLUTCH and confirmatory direct perturbation calculations in specific applications, e.g., fissionable and polyethylene reflectors

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Analysis of NCERC Critical Experiments with ENDF/B Nuclear Data Libraries

Nuclear data (ND) libraries are the backbone of the nuclear industry, as they are the collections of tabulated probabilities that define sub-atomic particle interactions with matter. In the areas of criticality safety and experiments, no evaluated nuclear data files (ENDF) are more important than those containing neutron cross section data. It is upon these files, and accompanying radiation transport codes, that practitioners are enabled to safely design both subcritical and critical systems. Likewise, in a symbiotic fashion, it is the same critical assemblies which are used primarily to validate that the cross sections are correct. The ENDF/B library, the United States’ national library maintained by the National Nuclear Data Center (NNDC) at Brookhaven National Laboratory (BNL), is soon releasing a new version, ENDF/B-VIII.1. Prior to the official release, several beta versions of the library were prepared and tested in simulation suites. The work presented here are results comparing the newest ENDF/B beta release (ENDF/B-VIII.1b3) and ENDF/B-VIII.0 with recent experiments done at the National Criticality Experiments Research Center (NCERC) as well as correlated experiments from the Los Alamos Critical Experiments Facility (LACEF). These NCERC and LACEF experiments were performed in part to provide validation for various cross sections that were identified as insufficient in the ENDF/B-VIII.0. In particular, lead, copper, fluorine, and chlorine, as well as the major actinides, were targeted from the last decade of critical experiments.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2020

Preliminary Criticality and Radiation Shielding Analysis for the Storage and Transfer of MARVEL Reactor Spent Nuclear Fuel

This report documents the results of preliminary nuclear criticality and radiation shielding assessments during transfer and dry storage of irradiated Microreactor Applications Research Validation and Evaluation (MARVEL) reactor fuel at Idaho National Laboratory (INL). The assessments focus on transfer casks and storage canisters that are currently in use at INL, which may be compatible with the irradiated MARVEL reactor fuel. The criticality assessments were performed using the radiation transport code MCNP6 with 37 MARVEL reactor fuel elements in various configurations and scenarios. All transfer and storage configurations under dry conditions were below the assumed criticality safety limit of 0.93. Some storage and transfer configurations under wet conditions exceeded the criticality safety limit. This suggests that the appropriate administrative and engineering controls, in addition to reducing the number of MARVEL reactor fuel elements per transfer cask or storage canister, can be expected to ensure criticality safety under all scenarios. The radiation shielding assessments were performed by generating conservative neutron and photon source terms using the ORIGEN module in the SCALE suite of codes. These source spectra were used to estimate the dose equivalent rates using the radiation transport code MCNP6, both on contact and 1 m away from the fuel and transfer casks. The maximum estimated dose equivalent rate of 37 unshielded MARVEL reactor fuel elements on contact is approximately 42000 R/hr. The maximum estimated dose equivalent rates on contact to the ATR transfer cask, HFEF-5 transfer cask, and high load charger were approximately 233 mR/hr, 171 mR/hr, and 201 mR/hr, respectively. This suggests that with the appropriate administrative and engineering controls, all three transfer casks analyzed can be expected to provide sufficient radiation shielding to workers during transfer of irradiated MARVEL reactor fuel. These calculations are performed to support the planning and strategy for the MARVEL project and will demonstrate the technical viability of the different configurations discussed and help identify where engineered or administrative controls may be necessary. A complete criticality safety analysis and radiation shielding analysis, including validation and contingency and accident analysis must be completed by licensed and authorized personnel before any transfer or storage of MARVEL reactor nuclear fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

EVALUATION OF HRA METHODOLOGIES FOR APPLICATION IN SDP WORK

This study critically evaluates human reliability analysis (HRA) methodologies applicable to regulatory probabilistic safety assessment (PSA) model, with a particular focus on their role in supporting the significance determination process (SDP) in nuclear safety assessment. Firstly, three widely utilized HRA methods – IDHEAS-ECA, SPAR-H, and ASEP/THERP – were qualitatively and quantitatively assessed. Qualitative assessments were conducted using attributes from the NEA/CSNI/R(2015)1 report, while quantitative evaluations employed regression and correlation analyses to compare predicted human error probabilities (HEPs) against empirical data. Results reveal distinct strengths, for example, IDHEAS-ECA’s robust predictive accuracy and K-HRA’s alignment with operational practices. In addition, dependency analysis and recovery analysis were critically evaluated. For dependency analysis, the methods’ handling of inter-task dependencies and their impact on HEPs were examined, while recovery analysis highlighted strategies for mitigating failure events. Furthermore, strategies were proposed to evaluate performance-shaping factors under conditions of reduced human performance, such as stress, fatigue, or cognitive overload, addressing specific challenges faced in SDP evaluations. Human errors from KINS’s operational performance information system event reports were evaluated as a case study. This study identifies gaps and provides actionable insights to ensure their validity and applicability in SDP HRA applications. This paper is a part of research conducted by KINS, and it should be noted that this result does not represent the regulatory position of KINS.

99 - GENERAL AND MISCELLANEOUS↗
NASA NTRS2008

Critical Initial Flaw Size Analysis

An independent assessment was conducted to determine the critical initial flaw size (CIFS) for the flange-to-skin weld in the Ares I-X Upper Stage Simulator (USS). The USS consists of several "tuna can" segments that are approximately 216 inches in diameter, 115 inches tall, and 0.5 inches thick. A 6 inch wide by 1 inch thick flange is welded to the skin and is used to fasten adjacent tuna cans. A schematic of a "tuna can" and the location of the flange-to-skin weld are shown in Figure 1. Gussets (shown in yellow in Figure 1) are welded to the skin and flange every 10 degrees around the circumference of the "tuna can". The flange-to-skin weld is a flux core butt weld with a fillet weld on the inside surface, as illustrated in Figure 2. The welding process may create loss of fusion defects in the weld that could develop into fatigue cracks and jeopardize the structural integrity of the Ares I-X vehicle. The CIFS analysis was conducted to determine the largest crack in the weld region that will not grow to failure within 4 lifetimes, as specified by NASA standard 5001 & 5019 [1].

Dawicke, David S.↗
DOE OSTI2022

Prediction of the energetic particle redistribution by an improved critical gradient model and analysis of the transport threshold

Based on the theory of critical gradient model (CGM) and following the simulation method proposed by Waltz et al., a combination of TGLFEP and EPtran code is employed to predict the energetic particle (EP) transport induced by Alfvén eigenmodes (AEs). To be consistent with the experiment, recent improvements to the simulation method include consideration of threshold evolution and orbit loss due to finite orbit width. The revised CGM is applied to simulate two DIII-D experimental discharges (#142111 and #153071). It well reproduces the experimental profiles with multiple unstable AEs and large-scale EP transport. Discharge #142111 had previously been simulated using a nonlinear MHD-kinetic code MEGA with a transport mechanism based on stochasticity induced by overlapping AE. By comparing the simulated EP profiles, we find that the AE transport threshold is approximated by both the MEGA nonlinear stability threshold and the proposed CGM threshold (error <5% for single n and <17% for multiple n simulation). Both of them are larger than the linear stability threshold of the most unstable AE mode by a quantity of the order of the flux needed to sustain EP transport by the background turbulence. Additionally, we have also applied the improved CGM to simulate the α particle redistribution for a China Fusion Engineering Test Reactor steady state scenario. Because of the clear separation between the AE unstable region and the loss cone, only a moderate α particle loss of ~9.6% is predicted.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗