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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 523 records · Page 29
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Advanced Fuels Campaign: Light Water Reactor Accident Tolerant Fuel Performance Metrics

The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. As a result, continual improvement of technology, including advanced materials and nuclear fuels, remains central to industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. In 2011, following the Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex, enhancing the accident tolerance of LWRs became a topic of serious discussion. As a result of direction from the U.S. Congress, the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) initiated an Accident Tolerant Fuel (ATF) Development program. The complex multiphysics behavior of LWR nuclear fuel makes defining specific material or design improvements difficult; as such, establishing qualitative attributes is critical to guide the design and development of fuels and cladding with enhanced accident tolerance. This report summarizes a common set of technical evaluation metrics to aid in the optimization and down selection of candidate designs. As used herein, “metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. Furthermore, this report describes a proposed technical evaluation methodology that can be applied to assess the ability of each concept to meet performance and safety goals relative to the current UO 2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed for lead test rod or lead test assembly insertion into a commercial reactor within the desired timeframe (by 2022).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

STATUS OF THE DEVELOPMENT OF IN-TANK/AT-TANK SEPARATIONS TECHNOLOGIES FOR FOR HIGH-LEVEL WASTE PROCESSING FOR THE U.S. DEPARTMENT OF ENERGY

Within the U.S. Department of Energy's (DOE) Office of Technology Innovation and Development, the Office of Waste Processing manages a research and development program related to the treatment and disposition of radioactive waste. At the Savannah River (South Carolina) and Hanford (Washington) Sites, approximately 90 million gallons of waste are distributed among 226 storage tanks (grouped or collocated in 'tank farms'). This waste may be considered to contain mixed and stratified high activity and low activity constituent waste liquids, salts and sludges that are collectively managed as high level waste (HLW). A large majority of these wastes and associated facilities are unique to the DOE, meaning many of the programs to treat these materials are 'first-of-a-kind' and unprecedented in scope and complexity. As a result, the technologies required to disposition these wastes must be developed from basic principles, or require significant re-engineering to adapt to DOE's specific applications. Of particular interest recently, the development of In-tank or At-Tank separation processes have the potential to treat waste with high returns on financial investment. The primary objective associated with In-Tank or At-Tank separation processes is to accelerate waste processing. Insertion of the technologies will (1) maximize available tank space to efficiently support permanent waste disposition including vitrification; (2) treat problematic waste prior to transfer to the primary processing facilities at either site (i.e., Hanford's Waste Treatment and Immobilization Plant (WTP) or Savannah River's Salt Waste Processing Facility (SWPF)); and (3) create a parallel treatment process to shorten the overall treatment duration. This paper will review the status of several of the R&D projects being developed by the U.S. DOE including insertion of the ion exchange (IX) technologies, such as Small Column Ion Exchange (SCIX) at Savannah River. This has the potential to align the salt and sludge processing life cycle, thereby reducing the Defense Waste Processing Facility (DWPF) mission by 7 years. Additionally at the Hanford site, problematic waste streams, such as high boehmite and phosphate wastes, could be treated prior to receipt by WTP and thus dramatically improve the capacity of the facility to process HLW. Treatment of boehmite by continuous sludge leaching (CSL) before receipt by WTP will dramatically reduce the process cycle time for the WTP pretreatment facility, while treatment of phosphate will significantly reduce the number of HLW borosilicate glass canisters produced at the WTP. These and other promising technologies will be discussed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Electric System Intra-hour Operation Simulator

ESIOS is a software program developed at Pacific Northwest National Laboratory (PNNL) that performs intra-hour dispatch and automatic generation control (AGC) simulations for electric power system frequency regulation and load/variable generation following. The program dispatches generation resources at minute interval to meet control performance requirements, while incorporating stochastic models of forecast errors and variability with generation, load, interchange and market behaviors. The simulator also contains an operator model that mimics manual actions to adjust resource dispatch and maintain system reserves. Besides simulating generation fleet intra-hour dispatch, ESIOS can also be used as a test platform for the design and verification of energy storage, demand response, and other technologies helping to accommodate variable generation.

Lu, Shuai↗
DOE OSTI2022

pyJSPEC - A Python Module for IBS and Electron Cooling Simulation

The intrabeam scattering is an important collective effect that can deteriorate the property of a high-intensity beam and electron cooling is a method to mitigate the IBS effect. JSPEC (JLab Simulation Package on Electron Cooling) is an open-source C++ program developed at Jefferson Lab, which simulates the evolution of the ion beam under the IBS and/or the electron cooling effect. The Python wrapper of the C++ code, pyJSPEC, for Python 3.x environment has been recently developed and released. It allows the users to run JSPEC simulations in a Python environment. It also makes it possible for JSPEC to collaborate with other accelerator and beam modeling programs as well as plentiful python tools in data visualization, optimization, machine learning, etc. In this paper, we will introduce the features of pyJSPEC and demonstrate how to use it with sample codes and numerical results.

Zhang, H.↗
DOE OSTI2017

Updated Liquid Secondary Waste Grout Formulation and Preliminary Waste Form Qualification

This report describes the results from liquid secondary waste grout (LSWG) formulation and cementitious waste form qualification tests performed by Pacific Northwest National Laboratory (PNNL) for Washington River Protection Solutions, LLC (WRPS). New formulations for preparing a cementitious waste form from a high-sulfate liquid secondary waste stream simulant, developed for Effluent Management Facility (EMF) process condensates merged with low activity waste (LAW) caustic scrubber, and the release of key constituents (e.g. 99 Tc and 129 I) from these monoliths were evaluated. This work supports a technology development program to address the technology needs for Hanford Site Effluent Treatment Facility (ETF) liquid secondary waste (LSW) solidification and supports future Direct Feed Low-Activity Waste (DFLAW) operations. High-priority activities included simulant development, LSWG formulation, and waste form qualification. The work contained within this report relates to waste form development and testing and does not directly support the 2017 integrated disposal facility (IDF) performance assessment (PA). However, this work contains valuable information for use in PA maintenance past FY17, and for future waste form development efforts. The provided data should be used by (i) cementitious waste form scientists to further understanding of cementitious dissolution behavior, (ii) IDF PA modelers who use quantified constituent leachability, effective diffusivity, and partitioning coefficients to advance PA modeling efforts, and (iii) the U.S. Department of Energy (DOE) contractors and decision makers as they assess the IDF PA program. The results obtained help fill existing data gaps, support final selection of a LSWG waste form, and improve the technical defensibility of long-term waste form performance estimates.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Updated Liquid Secondary Waste Grout Formulation and Preliminary Waste Form Qualification

This report describes the results from liquid secondary waste grout (LSWG) formulation and cementitious waste form qualification tests performed by Pacific Northwest National Laboratory (PNNL) for Washington River Protection Solutions, LLC (WRPS). New formulations for preparing a cementitious waste form from a high-sulfate liquid secondary waste stream simulant, developed for Effluent Management Facility (EMF) process condensates merged with low activity waste (LAW) caustic scrubber, and the release of key constituents (e.g. 99Tc and 129I) from these monoliths were evaluated. This work supports a technology development program to address the technology needs for Hanford Site Effluent Treatment Facility (ETF) liquid secondary waste (LSW) solidification and supports future Direct Feed Low-Activity Waste (DFLAW) operations. High-priority activities included simulant development, LSWG formulation, and waste form qualification. The work contained within this report relates to waste form development and testing and does not directly support the 2017 integrated disposal facility (IDF) performance assessment (PA). However, this work contains valuable information for use in PA maintenance past FY17, and for future waste form development efforts. The provided data should be used by (i) cementitious waste form scientists to further understanding of cementitious dissolution behavior, (ii) IDF PA modelers who use quantified constituent leachability, effective diffusivity, and partitioning coefficients to advance PA modeling efforts, and (iii) the U.S. Department of Energy (DOE) contractors and decision makers as they assess the IDF PA program. The results obtained help fill existing data gaps, support final selection of a LSWG waste form, and improve the technical defensibility of long-term waste form performance estimates.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

U.S. implementation of Q12{sup TM} zirconium alloy for improved stability of structural components

AREVA NP undertook the Q12{sup TM} research and development program to optimize the contribution to fuel assembly dimensional stability of structural components fabricated from ultra-low tin Zr-1Nb-Sn-Fe quaternary alloys. Because Q12{sup TM} is an evolutionary development of the current M5 alloy metallurgy, it benefits from the extensive industrial experience already gained with the latter. The Q12{sup TM} alloy is processed in the same way as M5, with slight increases in tin (Sn) and iron (Fe) contents. For structural components, these chemistry modifications provide higher irradiation creep strength and improved corrosion performance over Zircaloy-4. Fuel assemblies with quaternary alloy guide tubes are thus expected to increase robustness with respect to fuel assembly bow, improve growth characteristics, and maintain high burnup capability. The basic material properties have been determined through various in-core and out-of-core test programs. Components made from Q12{sup TM} have undergone irradiation in lead assemblies and in batch fuel for 13 Pressurized Water Reactor (PWR) units worldwide. Fuel with Q12{sup TM} guide tubes and grids has achieved a fuel assembly average burnup of 60 GWd/tU. Post irradiation examinations of these components have demonstrated that Q12{sup TM} performs in a stable, predictable manner. AREVA NP has submitted a topical report (TR) for Q12{sup TM} as a structural material to the United States Nuclear Regulatory Commission (US-NRC). This TR outlines the material behavior and performance of Q12{sup TM} for both guide tubes and grids. Once approved, implementation of Q12{sup TM} material throughout US PWR fuel designs will enhance dimensional stability of structural components. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Multifamily Housing Rehabilitation Process Improvements

Rea Ventures Group, LLC (Rea Ventures) partnered with Southface Energy Institute (Southface)—a member of the U.S. Department of Energy’s Partnership for Home Innovation Building America research team—to rehabilitate 418 low-income multifamily rental apartments located at 14 properties in Georgia (International Energy Conservation Code Climate Zones 2–4). These 22-year-old units with individual utility meters were arranged in row house or townhouse style. Rehabilitation plans were developed using a process prescribed by the U.S. Department of Agriculture (USDA) Rural Development program, which partially funded the building upgrades. The USDA is responsible for building, upgrading, and subsidizing housing in rural areas nationwide; this housing includes more than 14,000 existing multifamily housing developments. In 2012, more than $100 million in grants and loans were allocated for that purpose.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2016

SURROGATE MODEL DEVELOPMENT AND VALIDATION FOR RELIABILITY ANALYSIS OF REACTOR PRESSURE VESSELS

In nuclear light water reactors (LWRs), the reactor coolant, core and shroud are contained within a massive, thick walled steel vessel known as a reactor pressure vessel (RPV). Given the tremendous size of these structures, RPVs typically contain a large population of pre-existing flaws introduced in the manufacturing process. After many years of operation, irradiation-induced embrittlement makes these vessels increasingly susceptible to fracture initiation at the locations of the pre-existing flaws. Because of the uncertainty in the loading conditions, flaw characteristics and material properties, probabilistic methods are widely accepted and used in assessing RPV integrity. The Fracture Analysis of Vessels – Oak Ridge (FAVOR) computer program developed by researchers at Oak Ridge National Laboratory is widely used for this purpose. This program can be used in order to perform deterministic and probabilistic risk-informed analyses of the structural integrity of an RPV subjected to a range of thermal-hydraulic events. FAVOR uses a one-dimensional representation of the global response of the RPV, which is appropriate for the beltline region, which experiences the most embrittlement, and employs an influence coefficient technique to rapidly compute stress intensity factors for axis-aligned surface-breaking flaws. The Grizzly code is currently under development at Idaho National Laboratory (INL) to be used as a general multiphysics simulation tool to study a variety of degradation mechanisms in nuclear power plant components. The first application of Grizzly has been to study fracture in embrittled RPVs. Grizzly can be used to model the thermo-mechanical response of an RPV under transient conditions observed in a pressurized thermal shock (PTS) scenario. The global response of the vessel provides boundary conditions for local 3D models of the material in the vicinity of a flaw. Fracture domain integrals are computed to obtain stress intensity factors, which can in turn be used to assess whether a fracture would initiate at a pre-existing flaw. To use Grizzly for probabilistic analysis, it is necessary to have a way to rapidly evaluate stress intensity factors. To accomplish this goal, a reduced order model (ROM) has been developed to efficiently represent the behavior of a detailed 3D Grizzly model used to calculate fracture parameters. This approach uses the stress intensity factor influence coefficient method that has been used with great success in FAVOR. Instead of interpolating between tabulated solutions, as FAVOR does, the ROM approach uses a response surface methodology to compute fracture solutions based on a sampled set of results used to train the ROM. The main advantages of this approach are that the process of generating the training data can be fully automated, and the procedure can be readily used to consider more general flaw configurations. This paper demonstrates the procedure used to generate a ROM to rapidly compute stress intensity factors for axis-aligned flaws. The results from this procedure are in good agreement with those produced using the traditional influence coefficient interpolation procedure, which gives confidence in this method. This paves the way for applying this procedure for more general flaw configurations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

SU-E-J-100: Reconstruction of Prompt Gamma Ray Three Dimensional SPECT Image From Boron Neutron Capture Therapy(BNCT)

Purpose: Purpose of paper is to confirm the feasibility of acquisition of three dimensional single photon emission computed tomography (SPECT) image from boron neutron capture therapy (BNCT) using Monte Carlo simulation. Methods: In case of simulation, the pixelated SPECT detector, collimator and phantom were simulated using Monte Carlo n particle extended (MCNPX) simulation tool. A thermal neutron source (<1 eV) was used to react with the boron uptake region (BUR) in the phantom. Each geometry had a spherical pattern, and three different BURs (A, B and C region, density: 2.08 g/cm3) were located in the middle of the brain phantom. The data from 128 projections for each sorting process were used to achieve image reconstruction. The ordered subset expectation maximization (OSEM) reconstruction algorithm was used to obtain a tomographic image with eight subsets and five iterations. The receiver operating characteristic (ROC) curve analysis was used to evaluate the geometric accuracy of reconstructed image. Results: The OSEM image was compared with the original phantom pattern image. The area under the curve (AUC) was calculated as the gross area under each ROC curve. The three calculated AUC values were 0.738 (A region), 0.623 (B region), and 0.817 (C region). The differences between length of centers of two boron regions and distance of maximum count points were 0.3 cm, 1.6 cm and 1.4 cm. Conclusion: The possibility of extracting a 3D BNCT SPECT image was confirmed using the Monte Carlo simulation and OSEM algorithm. The prospects for obtaining an actual BNCT SPECT image were estimated from the quality of the simulated image and the simulation conditions. When multiple tumor region should be treated using the BNCT, a reasonable model to determine how many useful images can be obtained from the SPECT could be provided to the BNCT facilities. This research was supported by the Leading Foreign Research Institute Recruitment Program through the National Research Foundation of Korea (NRF) funded by the Ministry of Science, Information and Communication Technologies (ICT) and Future Planning (MSIP)(Grant No.200900420) and the Radiation Technology Research and Development program (Grant No.2013043498), Republic of Korea.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2015

Study on criticality control of fuel debris by Japan atomic energy agency to support regulation authority of Japan - 14571

From the viewpoint of safety regulation, criticality control of the fuel debris in the Fukushima Daiichi Nuclear Power Station would be a risk-informed control to mitigate consequences of criticality events, instead of a deterministic control to prevent such events. The Nuclear Regulation Authority of Japan has set up a research and development program to tackle this challenge. The Nuclear Safety Research Center of Japan Atomic Energy Agency, commissioned by the authority, has launched activities such as computations of criticality characteristics of the fuel debris, development of criticality risk assessment method, and preparation of criticality experiments to support them. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Nevada National Security Site: Site-Directed Research and Development (SDRD) Fiscal Year 2015 Annual Report

This report presents results of multiple research projects, new and ongoing, funded under the Site-Directed Research and Development Program for the Nevada National Security Site during federal fiscal year 2015. The Site's legacy capabilities in remote sensing combined with new paradigms for emergency response and consequence management help drive the need to develop advanced aerial sensor platforms. Likewise, dynamic materials science is a critical area of scientific research for which basic physics issues are still unresolved. New methods of characterizing materials in extreme states are vitally needed, and these efforts are paving the way with new knowledge. Projects selected in FY 2015 for the Exploratory Research portfolio exhibit a strong balance of NNSS mission relevance. Geoscience, seismology, and techniques for detecting underground nuclear events are still essential focus areas. Many of the project reports in the second major section of this annual report are ongoing continuations in multi-year lifecycles. Diagnostic techniques for stockpile and nuclear security science figured prominently as well, with a few key efforts coming to fruition, such as phase transition detection. In other areas, modeling efforts toward better understanding plasma focus physics has also started to pay dividends for major program needs.

42 ENGINEERING↗
DOE OSTI2010

Fiscal Year 2010 Well Plugging and Abandonment Reconfiguration Program Description Addendum

This work plan describes an addendum to the program for plugging and abandonment (P&A) or reconfiguration of groundwater monitoring wells at the Waste Isolation Pilot Plant (WIPP) facility for fiscal year 2010. The addendum adds work scope to the previously developed program plan (DOE/WIPP 08-3376). Over the years a number of wells have been installed to monitor temporal and spatial changes in groundwater elevation and chemistry. Many of the wells used for this purpose were cased with carbon steel pipe that has deteriorated to varying degrees. For some wells, the integrity has failed and they must be either reconfigured or completely plugged, according to regulatory requirements, to prevent possible commingling of water from different water bearing formations. Others are nearing the end of their useful life and/or are no longer needed for monitoring. This work plan presents the field activities and objectives, requisite documentation, project organization, and project team roles and responsibilities. Work described in this plan will be performed by a New Mexico well drilling contractor holding a valid and current well driller license. The term "Contractor" refers to the licensed well driller and /or registered drill rig supervisor assigned by the well drilling contractor, who will be directed by the WIPP Subcontract Technical Representative (STR) or designee at the drilling site. This Contractor will arrange lower-tier subcontracts, to include geophysics, tool rental, and waste containers.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Poster - 40: Treatment Verification of a 3D-printed Eye Phantom for Proton Therapy

Purpose: Ocular melanoma is a form of eye cancer which is often treated using proton therapy. The benefit of the steep proton dose gradient can only be leveraged for accurate patient eye alignment. A treatment-planning program was written to plan on a 3D-printed anatomical eye-phantom, which was then irradiated to demonstrate the feasibility of verifying in vivo dosimetry for proton therapy using PET imaging. Methods: A 3D CAD eye model with critical organs was designed and voxelized into the Monte-Carlo transport code FLUKA. Proton dose and PET isotope production were simulated for a treatment plan of a test tumour, generated by a 2D treatment-planning program developed using NumPy and proton range tables. Next, a plastic eye-phantom was 3D-printed from the CAD model, irradiated at the TRIUMF Proton Therapy facility, and imaged using a PET scanner. Results: The treatment-planning program prediction of the range setting and modulator wheel was verified in FLUKA to treat the tumour with at least 90% dose coverage for both tissue and plastic. An axial isotope distribution of the PET isotopes was simulated in FLUKA and converted to PET scan counts. Meanwhile, the 3D-printed eye-phantom successfully yielded a PET signal. Conclusions: The 2D treatment-planning program can predict required parameters to sufficiently treat an eye tumour, which was experimentally verified using commercial 3D-printing hardware to manufacture eye-phantoms. Comparison between the simulated and measured PET isotope distribution could provide a more realistic test of eye alignment, and a variation of the method using radiographic film is being developed.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2010

Marine Corps Base Camp Lejeune Shorebird Productivity Log 2008-2009

Critical military training and testing on lands along the nation’s coastal and estuarine shorelines are increasingly placed at risk because of development pressures in surrounding areas, impairments due to other anthropogenic disturbances, and increasing requirements for compliance with environmental regulations. The U.S. Department of Defense (DoD) intends to enhance and sustain its training and testing assets and to optimize its stewardship of natural resources through the development and application of an ecosystem-based management approach on DoD installations. To accomplish this goal, particularly for installations in estuarine/coastal environments, the Strategic Environmental Research and Development Program (SERDP) launched the Defense Coastal/Estuarine Research Program (DCERP) as a minimum 10-year effort at Marine Corps Base Camp Lejeune (MCBCL) in North Carolina. The overarching objectives of DCERP are to: (1) understand the effects of military training activities, infrastructure development, and other coastal military installations; (2) develop models, tools, and indicators to evaluate ecosystem health; and (3) recommend adaptive management strategies to sustain ecosystem natural resources within the context of an active military installation.This data package contains shorebird nest fate, morph measure, and band data from 2008-2009.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2013

Creep Rupture of Pressurized Alloy 617 Tubes

Creep behavior of Alloy 617 under multi-axial loading has been characterized at 950°C using thin walled tubes with internal gas pressure. Tubes were machined from Alloy 617 plate material to ensure comparability of material properties with extensive testing carried out in the VHTR Research and Development program. Properties of this alloy under multi-axial loading are of interest for elevated temperature design in advanced nuclear heat transfer systems. A constitutive model in the form of a Norton relationship was developed for Alloy 617 plate to allow development of a suitable creep specimen using finite element simulation. It was determined that proper description of the minimum creep rate at 750°C required a threshold stress formalism. The magnitude of the threshold stress was determined to be 75MPa for creep tests using the conditions examined. This threshold stress arises from strengthening by the γ’(Ni 3 Al,Ti) intermetallic phase; above this temperature this phase is unstable and Alloy 617 behaves as a solid solution.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Fabrication of UN microsphere kernels for Fully Ceramic Microencapsulated fuel

Fully Ceramic Microencapsulated (FCM) fuel is of interest to the U.S. Department of Energy's Fuel Cycle Research and Development program as an Accident Tolerant Fuel (ATF). The FCM design uses a TRISO (Tri-Structural Isotropic) particle with a ZrO{sub 2} surrogate kernel, embedded in a dense SiC matrix. TRISO technology, originally developed for High Temperature Gas Reactors, has an established high level of performance retaining fission products and the SiC matrix of an FCM compact shows good steam oxidation behavior. This work reports on a fabrication route for producing dense UN microspheres used as the kernel of the TRISO particles in FCM fuel. The UN microspheres are of high purity and density with enough porosity to accommodate volatile fission products. Increasing the density is nevertheless of interest potentially allowing for an additional adjustable property for optimizing the fuel kernel and TRISO particle characteristics. Current work is focused on investigating the effect of a feedstock microstructure on final product density. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗