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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 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

User’s Manual for Seal_Flux: A Seal Barrier Reduced-Order Model

This report provides a brief description on the use of the Seal_Flux computer program developed as part of the effort to quantify the risk of geologic storage of carbon dioxide (CO 2 ) under the U.S. Department of Energy’s (DOE) National Risk Assessment Partnership (NRAP). The Seal_Flux code simulates the flow of CO 2 through a low permeability rock horizon or seal formation overlying the storage reservoir into which CO 2 is injected. A two-phase, relative permeability approach with Darcy’s law is used for one-dimensional (1D) flow computations of CO 2 through the horizon in the vertical direction. The code also allows the simulation of time-dependent processes that can influence such flow.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2022

Initial Investigation of Coupled Electron-Photon Transport with SCEPTRE and RAPTURE (Rev.1.1)

The SCEPTRE1 code models coupled photon and electron transport in material media using discrete ordinates methods. The code can use different solvers in 1D and 3D in parallel, promising users the ability to run problems that are significant both in physics and computational scale. One capability that is perhaps unique at this time in the NNSA complex is SCEPTRE’s coupled photon– electron transport physics in a discrete ordinates approach. For example, incident x–rays can generate a significant electron population, whose transport can be simulated within the same code. In this context, fully–coupled means that photons can produce electrons and electrons can produce photons; uncoupled or no–coupling means photons do not produce electrons. Partial–coupling, an intermediate option in which photons can produce electrons but electrons do not produce photons, is not considered in the present work. The MCNP code can treat this physics with a Monte Carlo approach. MCNP should be viewed as complementary to SCEPTRE in both computational method (Monte Carlo versus discrete ordinates) as well as data libraries. SCEPTRE and MCNP should provide important checks on each other, especially from complementary modeling frameworks of a S n (discrete ordinates) mesh based method versus Monte Carlo particle based method. In order to learn more about SCEPTRE, and how it might inform engineering work at the Laboratory, we used it to carry out a set of initial studies. This was a series of 1D simulations of photon beams hitting Aluminum slabs. Our studies mainly used SCEPTRE directly, and made little use of the closely associated RAPTURE code (more on this below). The goals of these studies were first, to familiarize ourselves with operating the code, and second, to begin to assess the importance of coupled electron–photon transport. To support these studies, we began to develop a set of scripts to initialize inputs, run tools, process outputs, and plot results. An initial version of these scripts has been committed to the Lightning workflow repository. This memo relates the results of those studies.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Sierra/SD – Verification Test Manual – 5.22

Verification and validation (V&V) of scientific computing programs are important at Sandia National Labs due to the expanding role of computational simulation in managing the United States nuclear stockpile. The complexities of structural response calculations used to analyze physical problems, the varieties of codes applied to the calculations, and the importance of accurate predictions when assessing field conditions demand confidence in the consistency and accuracy of computer codes. Confidence in the accuracy of the predictions arising from computer simulations must ultimately be gained through verification and validation. The Sierra salinas structural dynamics analysis code, Sierra/SD, is used at the DOE Laboratories, and in several DOD projects. The roles of Sierra/SD in the qualification of weapon systems and components for normal and hostile environments throughout the Stockpile-to-Target Sequence include to, • Redesign weapon components. • Certify weapon components and systems for target environments such as hypersonic vehicles. • Certify that components will survive the thermal mechanical shock loads associated with hostile environments. • Evaluate current stockpile issues, including issues associated with uncertainty quantification. • Address many other problems that are encountered in stockpile management. The Sierra/SD verification plan is described, and an evolving set of key verification tests are described in detail. The verification tests ensure the correctness of the mathematics and numerical algorithms associated with functionality describing engineering phenomena. Development is in accordance with a set of tailored Software Quality Engineering (SQE) practices. SQE practices guide the overall verification and validation effort.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2025

Single-Bunch Instabilities Driven by Space Charge During Low-Energy Cooling at Injection in the EIC Hadron Storage Ring

This paper presents a simulation-based study of single-bunch dynamics at the injection energy of 23.8 GeV for protons in the EIC hadron storage ring, focusing on the impact of space-charge–driven instabilities. The analysis demonstrates that at this energy, the proton bunch experiences significant transverse space-charge forces, which can reduce the stability margin in the presence of the geometric and resistive wall impedance. Various collective effects were considered, with particular attention to the nonlinear nature of the transverse space charge. To stabilize the beam, high chromaticity and octupoles were introduced and their effects analyzed using the ELEGANT code. The results provide a quantitative assessment of the stability thresholds and offer guidance for the machine design and operational strategy at injection.

43 PARTICLE ACCELERATORS↗
DOE OSTI2021

Nuclear Data Sensitivity Study for the EBR-II Fast Reactor Benchmark Using SCALE with ENDF/B-VII.1 and ENDF/B-VIII.0

The EBR-II benchmark, which was recently included in the International Handbook of Evaluated Reactor Physics Benchmark Experiments, served as a basis for assessing the performance of the SCALE code system for fast reactor analyses. A reference SCALE model was developed based on the benchmark specifications. Great agreement was observed between the eigenvalue calculated with this SCALE model and the benchmark eigenvalue. To identify potential gaps and uncertainties of nuclear data for the simulation of various quantities of interest in fast spectrum systems, sensitivity and uncertainty analyses were performed for the eigenvalue, reactivity effects, and the radial power profile of EBR-II using the two most recent ENDF/B nuclear data library releases. While the nominal results are consistent between the calculations with the different libraries, the uncertainties due to nuclear data vary significantly. The major driver of observed uncertainties is the uncertainty of the 235 U ( n,γ ) reaction. Since the uncertainty of this reaction is significantly reduced in the ENDF/B-VIII.0 library compared to ENDF/B-VII.1, the obtained output uncertainties tend to be smaller in ENDF/B-VIII.0 calculations, although the decrease is partially compensated by increased uncertainties in 235 U fission and ν ¯ .

42 ENGINEERING↗
DOE OSTI2017

Uniaxial Ratcheting and Fatigue Failure of SA508-3 at Elevated Temperature

Ratcheting, describing the accumulation of cyclic plastic deformation under asymmetric cyclic stressing conditions, is a significant factor in the evaluation of fatigue life. SA508-3 ferritic stainless steel is a key material in the first loop of pressurized water reactor, often subjected to cyclic thermal and mechanical loadings with a temperature range of 25 deg. C to 350 deg. C. Therefore, the ratcheting behaviors of SA508-3 steel in service attract enough attentions. Both the American ASME and French RCC-M codes give specific descriptions about the assessment of the ratcheting behavior. However, it is difficult to predict the ratcheting strain since it is a secondary deformation superposed to the primary stress-strain response. Recent decades, the ratcheting behaviors of metals have been extensively studied experimentally and theoretically. Many of these researches mainly focused on the experiment observation of ratcheting behaviors within hundreds number of cycles and constitutive modeling without considering damage. A damage-coupled viscoplastic model was proposed by Kang et al. to simulate the whole-life ratcheting and predict the fatigue life of the tempered 42CrMo steel successfully, and thus it is worth to investigate the ratcheting-fatigue interaction for SA508-3 steel.

36 MATERIALS SCIENCE↗
DOE OSTI2018

DOE Residential Energy Code Field Study - Pennsylvania

This project consisted of three distinct phases – all focused on residential one- and two-family dwellings and townhouses. Phase 1 included a baseline field study with inspections and performance testing to assess construction practices with respect to the energy code. PNNL used the data to perform EUI and savings potential analyses. Phase 2 consisted of training, education, and outreach to building industry professionals with content based on the findings from Phase 1. Phase 3 was a post-program field study using the same methodology as Phase 1 and was intended to measure the impact of Phase 2. During Phase 1 of the project, PSD collected data from 171 job sites across eastern Pennsylvania, a territory bounded by PECO, PPL, and Met-Ed electric distribution territories, to reach the required 63 “home equivalents” deemed necessary by PNNL for a representative sample. Based on subsequent analysis, four key items were identified as having significant savings potential from improvement in compliance, including duct leakage, above-grade wall insulation (specifically insulation installation quality), foundation insulation, and high-efficacy lighting. About half the total savings potential is from duct leakage, so that was the primary focus. The goal of Phase 2 was to improve upon these key compliance weaknesses through an intensive training, education, and outreach program. PSD dubbed this effort the Penn Energy Codes Program. For Phase 3, PSD obtained a new sample plan from PNNL, obtained permit data from sample municipalities, and made substitutions where necessary. PSD field staff visited 165 job sites to collect the required 504 key observations. PSD submitted all data collection sheets to PNNL for their analysis to determine if there was a measurable change due to Phase 2 activities.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2019

System thermal-hydraulic modelling of the phénix dissymmetric test benchmark

Phénix is a French pool-type sodium-cooled prototype reactor; before the definitive shutdown, occurred in 2009, a final set of experimental tests are carried out in order to increase the knowledge on the operation and the safety aspect of the pool-type liquid metal-cooled reactors. One of the experiments was the Dissymmetric End-of-Life Test which was selected for the validation benchmark activity in the frame of SESAME project. The computer code validation plays a key role in the safety assessment of the innovative nuclear reactors and the Phénix dissymmetric test provides useful experimental data to verify the computer codes capability in the asymmetric thermal-hydraulic behaviour into a pool-type liquid metal-cooled reactor. This paper shows the comparison of the outcomes obtained with six different System Thermal-Hydraulic (STH) codes: RELAP5-3D©, SPECTRA, ATHLET, SAS4A/SASSYS-1, ASTEC-Na and CATHARE. The nodalization scheme of the reactor was individually achieved by the participants; during the development of the thermal-hydraulic model, the pool nodalization methodology had a special attention in order to investigate the capability of the STH codes to reproduce the dissymmetric effects which occur in each loop and into pools, caused by the azimuthal asymmetry of the boundary conditions. The modelling methodology of the participants is discussed and the main results are compared in this paper to obtain useful guide lines for the future modelling of innovative liquid metal pool-type reactors.

42 ENGINEERING↗
DOE OSTI2019

System Thermal-Hydraulic Modeling of the Phénix Dissymmetric Test Benchmark

Phénix is a French pool-type sodium-cooled prototype reactor; before the definitive shutdown, occurred in 2009, a final set of experimental tests are carried out in order to increase the knowledge on the operation and the safety aspect of the pool-type liquid metal-cooled reactors. One of the experiments was the Dissymmetric End-of-Life Test which was selected for the validation benchmark activity in the frame of SESAME project. The computer code validation plays a key role in the safety assessment of the innovative nuclear reactors and the Phénix dissymmetric test provides useful experimental data to verify the computer codes capability in the asymmetric thermal-hydraulic behaviour into a pool-type liquid metal-cooled reactor. This paper shows the comparison of the outcomes obtained with six different System Thermal-Hydraulic (STH) codes: RELAP5-3D©, SPECTRA, ATHLET, SAS4A/SASSYS-1, ASTEC-Na and CATHARE. The nodalization scheme of the reactor was individually achieved by the participants; during the development of the thermal-hydraulic model, the pool nodalization methodology had a special attention in order to investigate the capability of the STH codes to reproduce the dissymmetric effects which occurs in each loop and into pools, caused by the azimuthal asymmetry of the boundary conditions. The modelling methodology of the participants is discussed and the main results are compared in this paper to obtain useful guide lines for the future modelling of innovative liquid metal pool-type reactors.

Narcisi, V.↗
DOE OSTI2018

Analysis of Radiological Impacts to Terrestrial Biota in Support of Environmental Assessment for Use of DOE-Owned High-Assay Low-Enriched Uranium Stored at Idaho National Laboratory

There are many species of plants and animals (biota) on the INL Site and facilities such as INTEC and MFC have surrounding habitat supporting resident plant and animal populations. Consequently, radionuclide releases from the HALEAU processing facilities could pose a potential radiological impact to nearby flora and fauna populations. Therefore, the RESRAD-BIOTA (Version 1.8) computer code (http://resrad.evs.anl.gov/codes/resrad-biota/) was used to assess the impact of radiation exposures to terrestrial biota resulting from soil contaminated by airborne releases from the HALEAU processing facilities proposed for the INTEC and/or MFC complexes. The contaminated soil subsequently results in contamination in air and in different food sources used by biota. Both external radiation and internal radiation are considered in the dose evaluation.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2017

Cloud Computing for Complex Performance Codes.

This report describes the use of cloud computing services for running complex public domain performance assessment problems. The work consisted of two phases: Phase 1 was to demonstrate complex codes, on several differently configured servers, could run and compute trivial small scale problems in a commercial cloud infrastructure. Phase 2 focused on proving non-trivial large scale problems could be computed in the commercial cloud environment. The cloud computing effort was successfully applied using codes of interest to the geohydrology and nuclear waste disposal modeling community.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

LPG Component Leak Frequency Estimation

Liquefied petroleum gas (LPG) is used in heating, cooking, and as a vehicle fuel (called autogas). A safety risk assessment may be needed to assess potential hazard scenarios and inform the regulations, codes, and standards that apply to LPG facilities, such as autogas refueling facilities. The frequency of unintended releases in an LPG system is an important aspect of a system quantitative risk assessment. This report documents estimation of leakage frequencies for individual components of LPG systems. These frequencies are described using uncertainty distributions obtained with Bayesian statistical methods, generic data, and LPG data which were publicly available. There was a lack of LPG data in the literature, so frequencies for most components were developed with generic data and should be used cautiously; without additional information about component leak frequencies in LPG systems, it is not known whether these generic frequencies may be conservative or non-conservative.

02 PETROLEUM↗
DOE OSTI2020

MELCOR 2.2 Benchmarks of Peach Bottom NUREG/CR 7155 Uncertainty Analysis

The U.S. Nuclear Regulatory Commission (NRC) performed a first-of-a-kind uncertainty analysis (UA) of the accident progression, radiological releases, and offsite consequences for the State-of- the-Art Reactor Consequence Analyses (SOARCA) of an unmitigated long-term station blackout (LTSBO) severe accident scenario at the Peach Bottom Atomic Power Station. The objective of the UA was to evaluate the robustness of the SOARCA deterministic "best estimate results and conclusions documented in NUREG-1935, and to develop insight into the overall sensitivity of the SOARCA results to uncertainty in key modeling inputs. The study was completed in 2015 and documented in NUREG/CR-7155. Since 2015, two other SOARCA UAs were completed for two pressurized water reactor (PWR) plants. The PWR UAs incrementally updated the approach and methodology, including using the latest release of the MELCOR 2.2 computer code. There were also advances made in the state-of-the-art modeling related to NRC efforts using the Peach Bottom model in NUREG-2206, which provided the technical basis for the containment protection and release reduction rulemaking for boiling water reactors with Mark I and Mark II containments. This report documents the input model changes from the NUREG/CR-7155 study and performs a small number of reference calculations to assess the changes of the new computer code and the model input updates. The objective of the work is to verify whether the updated Peach Bottom MELCOR model and updated version of MELCOR support the conclusions formed in the Peach Bottom SOARCA UA by performing these representative calculations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Generation of group constants at GRS for the Rostov-2 benchmark

The OECD/NEA benchmark entitled 'Reactivity compensation of boron dilution by stepwise insertion of control rod cluster into the VVER-1000 core' aims at validating and assessing high fidelity multi-physics simulation code capabilities by comparison with the high-quality measurements performed at the nuclear power plant Rostov unit 2. The benchmark is divided into 2 phases: assembly wise and pin-by-pin resolution of the multi-physics problems. Multi-physics simulation requires the generation of an accurate parametrized few-group cross-section library, and so this benchmark offers an opportunity to validate the methodology for their generation. Based on the specifications of the benchmark, the core simulator KMACS developed at GRS was used to generate a library of cross-sections as well as the burnup distribution required to model the transient scenario of the benchmark at the assembly level. The assessment of the library is performed through comparison with measured values obtained at the initial state of the transient. Good agreement was obtained in terms of critical boron concentration, peaking factors and power profiles. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Combining RAVEN, RELAP5-3D, and PHISICS for Fuel Cycle and Core Design Analysis for New Cladding Criteria

The Nuclear Regulatory Commission (NRC) has considered revision of 10 CFR 50.46C rule to account for the effects of burn-up rate in future analysis of reactor accident scenarios so that safety margins may evolve with the operation and reloading of the reactor as dynamic limits. In order to find these limiting conditions, both cladding oxidation and maximum temperature must be cast as functions of fuel exposure. It is computationally intensive to run a plant model through a long duration operational transient to the time of fuel reload, and this must be repeated for each reload until the time of the accident scenario. Moreover, fuel reload patterns may differ in many ways so that calculations with many different core configurations must be considered. To perform such new analyses in a reasonable amount of computational time with good accuracy, Idaho National Laboratory (INL) has developed new multi-physics tools by combining existing codes and adding new capabilities. The Parallel Highly Innovative Simulation INL Code System (PHISICS) toolkit for neutronic and reactor physics is coupled with the Reactor Excursion and Leak Analysis Program (RELAP5-3D) for the Loss of Coolant Accident (LOCA) analysis and RAVEN for the Probabilistic Risk Assessment (PRA) and margin characterization analysis. Expansion of code capability was also necessary. For RELAP5-3D to process a single sequence of cores in a continuous run required modification of RELAP5-3D to restart a sequence of input decks, each with a slightly modified neutronics model or thermal-hydraulic flow region and culminating in an accident scenario. A new multi-deck input processing capability was developed and verified for this analysis. In conclusion, the combined RAVEN/PHISICS/RELAP5-3D tool is used to analyze a typical Pressurized Water Reactor (PWR).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Fitness For Service Assessment of a Corroded Heat Exchanger

Within the Fermi National Accelerator complex, there exist various water systems that support accelerator operations. One of these systems is extremely vital to the operation of the machine; that is the cooling system. The cooling system consists of nine relatively large heat exchangers that take untreated pond water and use it to cool the process fluid that further cools machine components. Over the 30 years these heat exchangers have been in operation, they have undergone significant material loss on the channels. This material loss, due to various forms of corrosion such as galvanic and microbiologically influenced corrosion (MIC) and possibly others, has deteriorated more than 80% of the nominal wall thickness of some of the exchangers and placed them in a questionable state. ASME FFS-1 (API 579) has been applied to address the condition of the heat exchangers due to their noncompliance with the governing code, BPVC Sec. VIII Div. 1. The assessments encompassed ASME FFS-1 parts 4: General Metal Loss and 9: Crack Like Flaw using level 1, 2, and 3 analysis techniques based on inspection data obtained by API 510 inspections. Level 1 and 2 assessments were deemed unfit for the corroded regions due to their location relative to a major structural discontinuity (channel to tube-sheet joint), so a level 3 analysis was conducted according to ASME Sec. VIII Div. 2 (design by analysis) rules for pressure vessels. Supplemental information included pond water tests to determine an accurate future corrosion allowance due to lacking inspection history. A leak before break (LBB) route was chosen to evaluate the possibility of leaking prior to the onset of failure. The analysis of one heat exchanger shows that the possibility the channel will develop a pinhole leak over 2.5 more years of operation should not be overlooked, but burst was unlikely from operation. The use of fracture mechanics show, that if a through-wall crack were to develop, it would not propagate further than the channel geometry and cause a leak not greater than 35 GPM. Using ASME Section XI Code Case N-705-1, allowing us to operate with a leak until the next outage given certain operating conditions and developing a leak mitigation procedure, this heat exchanger is deemed fit-for-service.

Humenik, Alex [Fermilab]↗
DOE OSTI2019

Model of temperature, evaporation, and productivity in elevated experimental algae raceways and comparison with commercial raceways

Elevated Experimental Raceways (EERs) have been used at many research sites in the United States to evaluate algal biomass productivity. This paper develops temperature and evaporation models for EERs and then evaluates whether the temperature profile and resultant productivity in an EER is representative of large-scale commercial raceways. Open water surface evaporation and temperature models with shading algorithms were programmed in Python and calibrated with temperature and evaporation data from the EERs in the Regional Algal Feedstock Testbed (RAFT) experiments at the University of Arizona. The final calibrated model for EERs was named the EERTEM (Elevated Experimental Raceway Temperature and Evaporation Model). The energy balance algorithms in the Biomass Assessment Tool (BAT model) were also added to the Python code and used to develop temperature simulations of standard commercial paddlewheel raceways. A comparison of BAT and EERTEM simulations indicated that EER temperature fluctuations are not representative of in ground commercial paddlewheel raceways, primarily due to the buffering effect of soil heat flux in commercial raceways. With the different temperature profiles, biomass productivities of three algae species in the EERs were compared to productivities in commercial raceways. Differences in productivity between EERTEM and BAT model temperature profiles were observed when the maximum daytime temperature of one raceway was in the optimal growth range but the temperature of the other raceway exceeded or was below the optimal growth range.

Khawam, George↗