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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 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

The 2019 Edition of the IRPhEP Handbook

The International Reactor Physics Experiment Evaluation Project (IRPhEP) [1] was established to preserve and evaluate integral reactor physics experiment data, including separate or special effects measurements for nuclear energy and technology applications. Numerous international experiments have been performed over the past several decades to support reactor operations, safety, measurements, and design. Those experiments represent a significant investment of infrastructure, cost, and expertise that, in many instances, cannot be as easily reproduced. They represent valuable resources of data for present and future research, and provide the basis for recording, development, and validation of methods [2]. The primary focus of the IRPhEP is to provide an extensively peer-reviewed set of reactor physics integral data suitable for reactor designers and safety analysts to not just validate the analytical tools necessary to design next-generation nuclear reactors, but to establish the safety basis for their operation. Contributors worldwide collaborate to evaluate and review selected benchmark experiments for inclusion in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook) [3]. More detailed discussion of the IRPhEP, including highlighted reactor physics benchmark evaluations, can be found in the November and December 2014 editions of Nuclear Science and Engineering [1]. Several new evaluations, and revisions to existing benchmarks have been prepared for inclusion in the 2019 edition of the IRPhEP Handbook (see Table I).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Coupling DAEM and CFD for simulating biomass fast pyrolysis in fluidized beds

We report results from computational simulations of an experimental, lab-scale bubbling bed biomass pyrolysis reactor that include a distributed activation energy model (DAEM) for the kinetics. In this study, we utilized multiphase computational fluid dynamics (CFD) to account for the turbulent hydrodynamics, and this was combined with the DAEM kinetics in a multi-component, multi-step reaction network. Our results indicate that it is possible to numerically integrate the coupled CFD–DAEM system without significantly increasing computational overhead. It is also clear, however, that reactor operating conditions, reaction kinetics, and multiphase flow dynamics all have major impacts on the pyrolysis products exiting the reactor. We find that, with the same pre-exponential factors and mean activation energies, inclusion of distributed activation energies in the kinetics can shift the predicted average value of the exit vapor-phase tar flux and its statistical distribution, compared to single-valued activation-energy kinetics. Perhaps the most interesting observed trend is that increasing the diversity of the DAEM activation energies appears to increase the mean tar yield, all else being equal. As a result, these findings imply that accurate resolution of the reaction activation energy distributions will be important for optimizing biomass pyrolysis processes.

09 BIOMASS FUELS↗
DOE OSTI2017

Accelerated materials evaluation for nuclear applications

This report addresses the opportunities and complexities of using materials test reactors with high neutron fluxes to perform accelerated studies of material aging in power reactors operating at lower neutron fluxes and with different neutron flux spectra. Radiation damage and gas production in different reactors have been compared using the code, SPECTER. This code provides a common standard from which to compare neutron damage data generated by different research groups using a variety of reactors.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Corrigendum to “Accelerated materials evaluation for nuclear applications” [J. Nucl. Mater. 488 (2017) 46–62]

The original article addresses the opportunities and complexities of using materials test reactors with high neutron fluxes to perform accelerated studies of material aging in power reactors operating at lower neutron fluxes and with different neutron flux spectra. Radiation damage and gas production in different reactors have been compared using the code, SPECTER. This code provides a common standard from which to compare neutron damage data generated by different research groups using a variety of reactors. This Corrigendum identifies a few typographical errors. Tables 2 and 3 are included in revised form.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

POLCA8 - modelling of cross section variations inside hexagonal assemblies

This paper presents the POLCA8 approach for modelling non-constant cross section distributions inside hexagonal fuel assemblies. The multigroup diffusion equation is modified to account for intranodal cross section variations. The obtained equation is solved in a node-wise manner based on the Fourier expansion method. As a result of varying cross sections, the solution includes a particular part additionally to the homogeneous one. A method for obtaining the particular solution is derived. Numerical tests on a VVER-1000 core are presented showing the impact of cross-section variations to some key parameters for reactor operation. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

PTT System Performance Evaluation in 3-D Imaging of Calibrated Defects: Pulsed Thermal Tomography Nondestructive Examination of Additively Manufactured Reactor Materials and Components

Research activity in this project is aimed at developing pulsed thermal tomography (PTT) architecture and algorithms for in-service non-destructive examination (NDE) of additively manufactured (AM) reactor components and materials. Recent advances in AM technology could allow cost-efficient fabrication of parts for operating reactors. However, because of stringent safety requirements, long-term performance of AM reactor components needs to be investigated before AM is widely accepted. Because of the complex shapes and significant surface roughness of components due to layer-by-layer welding process in AM, conventional methods might not be useful for in-service NDE of AM components. PTT is ideally suited for in-service NDE of AM metallic components because the method is non-contact, and offers high resolution #-D imaging of material flaws. The objective of this report is to provide initial evaluate of PTT performance sin detecting calibrated flaws in reactor structural materials. Preliminary COMSOL models were developed to conduct super computing simulations of PTT. In the experimental studies, high strength Stainless Steel 316 and Inconel 718 alloys were considered, as well as lower grade Stainless Steel 304, Nickel 200, and Hastelloy C276. Specimens investigated in this report consisted of approximately 1/4in-thick plates made out of these alloys using conventional manufacturing methods. The calibrated defects were created in the form of flat bottom holes (FBH) drilled in metallic plates. The diameters of FBH's varied from 1mm to 8mm, and their depths below the plate flat surface varied between 1mm and 6mm. The size of the smallest FBH was limited to 1mm because conventional mechanic drills were used for creating the holes. PTT imaging results have shown that 1mm-diameter FBH located 1mm and 2mm below the surface were detectable. Larger size FBH were detectable at greater depth. For example, 6mm-diameter FBH could be detected at 8mm depth. Image contrast varied slightly between the specimens, with the best reconstructions obtained in SS316 and C276 plates. In addition, a 2/3in-thick Inconel 718 nozzle plate produced with additive manufacturing method was imaged with PTT. It was shown that PTT can scan through the plate in approximately 20s. Several modes of 3-D data visualization were explored, including using ImageJ and MATLAB software packages.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Irradiation effects in Generation IV nuclear reactor materials

Generation IV reactor structural materials will be exposed to high doses and temperatures during reactor operation that may lead to irradiation-induced degradation. This degradation will differ from that seen in light water reactors and therefore understanding mechanisms controlling material performance during irradiation is critical for evaluating viability of Generation IV nuclear reactor concepts. This chapter discusses irradiation effects and microstructural changes that affect mechanical properties and dimensional stability of Generation IV reactor materials.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Progress on Grizzly Development for Reactor Pressure Vessels and Reinforced Concrete Structures

The Grizzly code has been under development for the U.S. Department of Energy’s Light Water Reactor Sustainability program to provide predictive tools for the evolution of material degradation in critical light water reactor structural components due to long-term exposure to the environmental conditions of normal reactor operation, and for the effects of this degradation on the safety of these components. This development has primarily targeted reactor pressure vessels and reinforced concrete structures. Work performed during Fiscal Year 2019 added important features for both of these types of structures. For probabilistic fracture mechanics analysis of reactor pressure vessels, capabilities were added to account for the effects of warm prestressing and residual stresses, as well as to model crack initiation, growth, and arrest. For modeling microstructure evolution in reactor pressure vessels, a cluster dynamics model for evolution of Mn-Ni-Si phases has been implemented. For modeling degraded reinforced concrete structures, capabilities to consider the combined effects of damage and creep in the concrete constitutive model and model nonlinear behavior of reinforcing bars has been added. These newly-developed features fill important gaps in Grizzly’s feature set for engineering-scale analysis of degraded reactor pressure vessels and reinforced concrete structures. Grizzly development in both of these areas has reached a point where it is ready for extensive testing on relevant engineering problems.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Compendium of HFIR and REDC Infrasound and Low-Frequency Acoustic Emissions

This document is a compilation of infrasound and low-frequency acoustic signatures that characterize select nuclear reactor operations and hot-cell operations observed at Oak Ridge National Laboratory (ORNL) in fiscal year 2018. Also included here are the results of digital signal analyses describing these signatures. This effort provides insight about the feasibility of using acoustic signatures for local and remote-sensing and identification of nuclear fuel-cycle operations. The signatures included here, in this first year of on-site observations at ORNL, will become part of the broader basis set of signatures collected by researchers involved in the Multi-Informatics for Nuclear Operations Scenarios (MINOS) venture. The effort by MINOS researchers is to gain a deeper knowledge base of nuclear facility operations which may be used to improve nuclear facility monitoring capabilities. This compendium comprises signatures developed from sound signals that were measured at ORNL's High Flux Isotope Reactor and the Radiochemical Engineering Development Center during the week of June 18, 2018. Most facility operations were performed specifically for this campaign, although some measurements were obtained in the course of normal operations. The knowledge gained from developing the signatures included herein will provide a better understanding that will be useful for upcoming measurement campaigns; as additional data becomes available it will be included in newer versions of this compendium.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2018

Compendium of HFIR and REDC Infrasound and Low-Frequency Acoustic Emissions

This document is a compilation of infrasound and low-frequency acoustic signatures that characterize select nuclear reactor operations and hot-cell operations observed at Oak Ridge National Laboratory (ORNL) in fiscal year 2018. Also included here are the results of digital signal analyses describing these signatures. This effort provides insight about the feasibility of using acoustic signatures for local and remote-sensing and identification of nuclear fuel-cycle operations. The signatures included here, in this first year of on-site observations at ORNL, will become part of the broader basis set of signatures collected by researchers involved in the Multi-Informatics for Nuclear Operations Scenarios (MINOS) venture. The effort by MINOS researchers is to gain a deeper knowledge base of nuclear facility operations which may be used to improve nuclear facility monitoring capabilities. This compendium comprises signatures developed from sound signals that were measured at ORNL's High Flux Isotope Reactor and the Radiochemical Engineering Development Center during the week of June 18, 2018. Most facility operations were performed specifically for this campaign, although some measurements were obtained in the course of normal operations. The knowledge gained from developing the signatures included herein will provide a better understanding that will be useful for upcoming measurement campaigns; as additional data becomes available it will be included in newer versions of this compendium.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Mechanical Testing of Hydrided Fuel Cladding

The U.S. Department of Energy Office of Nuclear Energy, Office of Fuel Cycle Technology established the Used Fuel Disposition Campaign (UFDC) to conduct research and development activities related to storage, transportation, and disposal of used nuclear fuel (UNF) and high-level radioactive waste. Within the UFDC, the storage and transportation task has been created to address issues of extended storage and transportation. One gap in knowledge identified as high priority by UFDC includes hydrogen effects in zirconium alloy cladding and reorientation of cladding hydrides. During reactor operations, hydrogen is generated from the oxidation of the surface of zirconium alloy materials in the coolant of light water reactors can be absorbed into the cladding and form zirconium hydrides. At high temperatures and hoop stresses, these normally circumferential oriented hydrides in high burnup UNF (HB-UNF) may reorient in the radial direction and affect the mechanical properties of the cladding. Understanding these changes in mechanical properties will aid in analyzing how UNF will perform during storage and transportation. Separate effects testing to measure the mechanical properties of un-hydrided, pre-hydrided, and irradiated HB-UNF is currently underway at Pacific Northwest National Laboratory (PNNL). In order to measure these mechanical properties, staff at PNNL are employing tensile, compression, and burst testing. Systems have been developed to pre-hydride cladding and reorient cladding hydrides. The development of these systems and the testing methods are described herein.

Cladding, Zirconium, Hydrides, Mechanical Testing↗
DOE OSTI2020

Assessment of Near-Term Fuel Screening and Qualification Needs for Nuclear Thermal Propulsion Systems

Nuclear thermal propulsion (NTP) is an in-space propulsion technology capable of both high specific impulse (850–1000 s) and thrust (44–1112 kN), which can help reduce trip times for crewed missions beyond low Earth orbit. NTP technology has been demonstrated during historic programs. Over 20 ground test reactor experiments were performed, which demonstrated the prototypic reactor operations, during the Nuclear Engine for Rocket Vehicle Application (NERVA)/Rover program (1955–1972). Although historical programs have shown that NTP is a viable in-space propulsion technology, developing NTP in modern programs is contingent on the development and qualification of ultrahigh-temperature nuclear fuel technologies that can withstand engine operating conditions. In historical NTP development programs such as NERVA/Rover, prototypic reactor/engine schemes were ground tested to assess the overall system feasibility and to qualify the reactor fuel forms for eventual flight systems. Although this approach is effective to verify fuel performance under prototypic conditions, relying solely on full-scale NTP reactor tests as the pathway for verifying or qualifying fuel is inefficient and cost prohibitive today. Additionally, modern nuclear licensing requirements state that before test reactor approval, reactor components and fuel elements should be qualified via non-nuclear (out-of-pile) and nuclear (in-pile) testing under representative operating conditions. Using this methodology, fuel matures as production scale fabrication methods are established, and as produced fuel performance is demonstrated. Here, this paper provides an overview of historical approaches to NTP fuel performance maturation, including fuel screening and qualification needs, and provides insight for establishing an efficient testing paradigm that can be implemented to rapidly and affordably develop NTP fuel forms for eventual qualification.

42 ENGINEERING↗
DOE OSTI2021

Critical Assemblies: Dragon Burst Assembly and Solution Assemblies

This work reviews the historical literature associated with the Dragon experiment and water boiler reactors operated at Los Alamos during the Manhattan Project. Frisch’s invited talk given at the American Nuclear Society’s Fast Burst Reactor Conference held at the University of New Mexico in Albuquerque, New Mexico, in 1969 is quoted. From the literature review, basic models for the Dragon experiment and for a water boiler–type assembly (aqueous homogeneous reactor) were created that can be used for conducting multiphysics simulations for criticality excursion studies. This methodology utilizes the coupled neutronic-hydrodynamic method to perform a time-dependent dynamic simulation of a criticality excursion. MCNP® was utilized to calculate important nuclear kinetic parameters that were incorporated into the models. Simulation results compare reasonably well with historic data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Used Fuel Disposition Campaign: Embrittlement and DBTT of High-Burnup PWR Fuel Cladding Alloys

Structural analyses of high-burnup (HBU) fuel require cladding mechanical properties and failure limits to assess fuel behavior during long-term dry-cask storage and transportation. Pre-storage drying-transfer operations and early stage storage subject cladding to higher temperatures and pressure-induced tensile hoop stresses relative to in-reactor operation and pool storage. Under these conditions, radial hydrides may precipitate during slow cooling and provide an additional embrittlement mechanism as the cladding temperature decreases below the ductile-to-brittle transition temperature (DBTT). On the basis of previous test results, susceptibility to radial-hydride precipitation depends on cladding material and microstructure, hydrogen content (CH), and pre-drying hydride distribution, as well as peak hoop stresses and temperatures during drying operations and storage. Susceptibility to embrittlement depends on the extent of radial-hydride precipitation, the number density of circumferential hydrides, and the thickness of the outer-surface hydride rim. The goal of the testing program is to determine drying-storage conditions for which the cladding remains ductile at ambient temperature, in other words, with the DBTT ≤20°C.

Billone, M. C.↗
DOE OSTI2023

National Reactor Innovation Center (NRIC) Infrastructure Integration Plan

The NRIC Infrastructure Integration Plan serves as a roadmap for integrating future test reactors at the Idaho National Laboratory’s (INL’s) desert Site into core infrastructure and services required for reactor operations. This document is not intended to serve as a requirements document, nor does it list all infrastructure requirements for each reactor planned for testing at the Site. It represents an overview of the most likely infrastructure and services that will be required by some or all the test reactors, as well as what infrastructure and services exist at the Site that have the potential to support and enable NRIC test reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Preparation of Simulated LBL Defects for Round Robin Experiment

A critical characteristic of the TRISO fuel design is its ability to retain fission products. During reactor operation, the TRISO layers act as barriers to release of fission products not stabilized in the kernel. Each component of the TRISO particle and compact construction plays a unique role in retaining select fission products, and layer performance is often interrelated. The IPyC, SiC, and OPyC layers are barriers to the release of fission product gases such as Kr and Xe. The SiC layer provides the primary barrier to release of metallic fission products not retained in the kernel, as transport across the SiC layer is rate limiting due to the greater permeability of the IPyC and OPyC layers to many metallic fission products. These attributes allow intact TRISO coatings to successfully retain most fission products released from the kernel, with the majority of released fission products during operation being due to defective, damaged, or failed coatings. This dominant release of fission products from compromised particles contributes to the overall source term in reactor; causing safety and maintenance concerns and limiting the lifetime of the fuel. Under these considerations, an understanding of the nature and frequency of compromised particles is an important part of predicting the expected fission product release and ensuring safe and efficient operation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Driving towards cost-competitive biofuels through catalytic fast pyrolysis by rethinking catalyst selection and reactor configuration

Catalytic fast pyrolysis (CFP) has emerged as an attractive process for the conversion of lignocellulosic biomass into renewable fuels and products. Considerable research and development has focused on using circulating-bed reactors with zeolite catalysts (e.g., HZSM-5) for CFP because of their propensity to form gasoline-range aromatic hydrocarbons. However, the high selectivity for aromatics comes at the expense of low carbon yield, a key economic driver for this process. In this contribution, we evaluate non-zeolite catalysts in a fixed-bed reactor configuration for an integrated CFP process to produce fuel blendstocks from lignocellulosic biomass. These experimental efforts are coupled with technoeconomic analysis (TEA) to benchmark the process and guide research and development activities to minimize costs. The results indicate that CFP bio-oil can be produced from pine with improved yield by using a bifunctional metal-acid 2 wt% Pt/TiO 2 catalyst in a fixed-bed reactor operated with co-fed H 2 at near atmospheric pressure, as compared to H-ZSM5 in a circulating-bed reactor. The Pt/TiO 2 catalyst exhibited good stability over 13 reaction-regeneration cycles with no evidence of irreversible deactivation. The CFP bio-oil was continuously hydrotreated for 140 h time-on-stream using a single-stage system with 84 wt% of the hydrotreated product having a boiling point in the gasoline and distillate range. This integrated biomass-to-blendstock process was determined to exhibit an energy efficiency of 50% and a carbon efficiency of 38%, based on the experimental results and process modelling. TEA of the integrated process revealed a modelled minimum fuel selling price (MFSP) of $4.34 per gasoline gallon equivalent (GGE), which represents a cost reduction of $0.85 GGE -1 compared to values reported for CFP with a zeolite catalyst. TEA also indicated that catalyst cost was a significant factor influencing the MFSP, which informed additional CFP experiments in which lower-cost Mo2C and high-dispersion 0.5 wt% Pt/TiO 2 catalysts were synthesized and evaluated. These materials demonstrated CFP carbon yield and oil oxygen content similar to those of the 2 wt% Pt/TiO 2 catalyst, offering proof-of-concept that the lower-cost catalysts can be effective for CFP and providing a route to reduce the modelled MFSP to $3.86–3.91 GGE -1 . This report links foundational science and applied engineering to demonstrate the potential of fixed-bed CFP and highlights the impact of coupled TEA to guide research activities towards cost reductions.

09 BIOMASS FUELS↗