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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 1,243 records · Page 69
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

INITIAL ANALYSIS OF TRANSIENT POWER TIME LAG DUE TO HETEROGENEITY WITHIN THE TREAT FUEL MATRIX.

The topic Nuclear Safety encompasses a broad spectrum of focal areas within the nuclear industry; one specific aspect centers on the performance and integrity of nuclear fuel during a reactivity insertion accident (RIA). This specific accident has proven to be fundamentally difficult to theoretically characterize due to the numerous empirically driven characteristics that quantify the fuel and reactor performance. The Transient Reactor Test (TREAT) facility was designed and operated to better understand fuel behavior under extreme (i.e. accident) conditions; it was shutdown in 1994. Recently, efforts have been underway to commission the TREAT facility to continue testing of advanced accident tolerant fuels (i.e. recently developed fuel concepts). To aid in the restart effort, new simulation tools are being used to investigate the behavior of nuclear fuels during facility’s transient events. This study focuses specifically on the characterizing modeled effects of fuel particles within the fuel matrix of the TREAT. The objective of this study was to (1) identify the impact of modeled heterogeneity within the fuel matrix during a transient event, and (2) demonstrate acceptable modeling processes for the purpose of TREAT safety analyses, specific to fuel matrix and particle size. Hypothetically, a fuel that is dominantly heterogeneous will demonstrate a clearly different temporal heating response to that of a modeled homogeneous fuel. This time difference is a result of the uniqueness of the thermal diffusivity within the fuel particle and fuel matrix. Using MOOSE/BISON to simulate the temperature time-lag effect of fuel particle diameter during a transient event, a comparison of the average graphite moderator temperature surrounding a spherical particle of fuel was made for both types of fuel simulations. This comparison showed that at a given time and with a specific fuel particle diameter, the fuel particle (heterogeneous) simulation and the homogeneous simulation were related by a multiplier relative to the average moderator temperature. As time increases the multiplier is comparable to the factor found in a previous analytical study from literature. The implementation of this multiplier and the method of analysis may be employed to remove assumptions and increase fidelity for future research on the effect of fuel particles during transient events.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Systems for the Intermodal Routing of Spent Nuclear Fuel

The safe and secure movement of spent nuclear fuel from shutdown and active reactor facilities to intermediate or long term storage sites may, in some instances, require the use of several modes of transportation to accomplish the move. To that end, a fully operable multi-modal routing system is being developed within Oak Ridge National Laboratory s (ORNL) WebTRAGIS (Transportation Routing Analysis Geographic Information System). This study aims to provide an overview of multi-modal routing, the existing state of the TRAGIS networks, the source data needs, and the requirements for developing structural relationships between various modes to create a suitable system for modeling the transport of spent nuclear fuel via a multimodal network. Modern transportation systems are comprised of interconnected, yet separate, modal networks. Efficient transportation networks rely upon the smooth transfer of cargoes at junction points that serve as connectors between modes. A key logistical impediment to the shipment of spent nuclear fuel is the absence of identified or designated transfer locations between transport modes. Understanding the potential network impacts on intermodal transportation of spent nuclear fuel is vital for planning transportation routes from origin to destination. By identifying key locations where modes intersect, routing decisions can be made to prioritize cost savings, optimize transport times and minimize potential risks to the population and environment. In order to facilitate such a process, ORNL began the development of a base intermodal network and associated routing code. The network was developed using previous intermodal networks and information from publicly available data sources to construct a database of potential intermodal transfer locations with likely capability to handle spent nuclear fuel casks. The coding development focused on modifying the existing WebTRAGIS routing code to accommodate intermodal transfers and the selection of prioritization constraints and modifiers to determine route selection. The limitations of the current model and future directions for development are discussed, including the current state of information on possible intermodal transfer locations for spent fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Computerized Operator Support System – Phase II Development

A computerized operator support system (COSS) prototype for nuclear control room process control is proposed and discussed. The COSS aids operators in addressing rapid plant upsets that would otherwise result in the shutdown of the power plant and interrupt electrical power generation, representing significant costs to the owning utility. In its current stage of development the prototype demonstrates four advanced functions operators can use to more efficiently monitor and control the plant. These advanced functions consist of: (1) a synthesized and intuitive high level overview display of system components and interrelations, (2) an enthalpy-based mathematical chemical and volume control system (CVCS) model to detect and diagnose component failures, (3) recommended strategies to mitigate component failure effects and return the plant back to pre-fault status, and (4) computer-based procedures to walk the operator through the recommended mitigation actions. The COSS was demonstrated to a group of operators and their feedback was collected. The operators responded positively to the COSS capabilities and features and indicated the system would be an effective operator aid. The operators also suggested several additional features and capabilities for the next iteration of development. Future versions of the COSS prototype will include additional plant systems, flexible computer-based procedure presentation formats, and support for simultaneous component fault diagnosis and dual fault synergistic mitigation action strategies to more efficiently arrest any plant upsets.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

IMPLEMENTING THE NFPA 805 PROCESS: Observations of a Technical Reviewer

In June 2004 the U.S. Nuclear Regulatory Commission (NRC) amended its fire protection requirements to permit existing nuclear power reactor licensees to voluntarily adopt fire protection requirements contained in National Fire Protection Association (NFPA) Standard 805. NFPA 805 is a performance-based standard for nuclear power plant fire protection that is an alternative to the deterministic, prescriptive fire protection requirements, such as 10 CFR 50 Appendix R, that was issued in 1980. One aspect of implementing NFPA 805 is that the licensee adopts the performance goals, objectives, and criteria for nuclear safety specified in the Standard. These goals, objectives, and criteria can be met through the implementation of deterministic approaches or performance-based approaches, including engineering analyses, probabilistic risk assessment, and fire modeling. Licensees voluntarily adopting the fire protection requirements in NFPA 805 must submit a license amendment request (LAR) to the NRC. The LAR provides the new proposed fire protection licensing basis, including the methodology and results of required evaluations and analyses that show how the NFPA 805 performance criteria are met. As of August 2014, licensees have submitted LARs for 26 nuclear power plants, representing 42 nuclear reactor units. Of these, 7 nuclear power plants, representing 10 nuclear reactor units, have been issued a safety evaluation (SE) by the NRC approving transition of their fire protection licensing basis to one that complies with NFPA 805. Pacific Northwest National Laboratory (PNNL) supports the NRC staff’s technical review of the LARs in the areas of fundamental fire protection, safe shutdown analysis, and Probabilistic Risk Assessment (PRA). PNNL, of course, cannot speak for the nuclear industry and its choice of implementation strategies or the NRC staff’s assessment of the approaches being taken to adopt NFPA 805. However, as a reviewer of the technical details of these submittals, PNNL is in a position to observe the array of implementation tactics taken in these submittals, and observe different ways licensees are making the NFPA 805 process work. For example, we see differences in how fire areas are being transitioned, the kinds of plant modifications being implemented, the changes being made to plant procedures, the number and types of recovery actions being credited, and the kinds and extent of detailed modeling being performed in support of the Fire PRAs. As a caveat, we note that it is probably too early to comment on the overall success or limitations of the NFPA 805 process or provide lessons learned for the future. Furthermore, it is not our intention to endorse any particular approach taken in a submittal over another or to critique the industry or the regulator. Rather our goal in this paper is to summarize a set of interesting and useful differences across submittals that may provide context for further future discussions about what we (i.e., reviewers, industry, and regulators) have learned in being part of the NFPA process; and how to best use that information to inform future NFPA 805 activities or other risk-informed endeavors.

NFPA 805, PRA, PSA, probabilistic risk assessment,↗
DOE OSTI2011

Immediate Deployment of Waste Energy Technologies at Multiple Sites

Verso Paper Corp. implemented a portfolio of 13 commercially available proven industrial technologies each exceeding 30% minimum threshold efficiency and at least 25% efficiency increase. These sub-projects are a direct result of a grant received from the Department of Energy (DOE) through its FOA 0000044 (Deployment of Combined Heat and Power (CHP) Systems, District Energy Systems, Waste Energy Recovery Systems, and Efficient Industrial Equipment), which was funded by the American Recovery Act. These were installed at 3 sites in 2 states and are helping to reduce Verso costs, making the facilities more competitive. This created approximately 100 construction jobs (FTE’s) and reduced impacted Verso facilities’ expense budgets. These sub-projects were deployed at Verso paper mills located in Jay, Maine, Bucksport, Maine, and Sartell, Minnesota. The paper mills are the economic engines of the rural communities in which these mills are located. Reinvestment in waste energy recovery capital improvements is providing a stimulus to help maintain domestic jobs and to competitively position the US pulp and paper industry with rising energy costs. Energy efficiency improvements are also providing a positive environmental impact by reducing greenhouse gas emissions, the quantity of wastewater treated and discharged, and fossil fuel demand. As a result of these projects, when fully operating, Verso realized a total of approximately 1.5 TBtu/Year reduction in overall energy consumption, which is 119% of the project objectives. Note that three paper machines have since been permanently curtailed. However even with these shutdowns, the company still met its energy objectives. Note also that the Sartell mill’s paper machine is down due to a recent fire which damaged the mill’s electrical infrastructure (the company has not decided on the mill’s future).

distributed energy, combined heat and power, CHP, ↗
DOE OSTI2012

Data Center Energy Efficiency Measurement Assessment Kit Guide and Specification

A portable and temporary wireless mesh assessment kit can be used to speed up and reduce the costs of a data center energy use assessment and overcome the issues with respect to shutdowns. The assessment kit is comprised of temperature, relative humidity, and pressure sensors. Also included are power meters that can be installed on computer room air conditioners (CRACs) without intrusive interruption of data center operations. The assessment kit produces data required for a detailed energy assessment of the data center.

Data, Center, Energy, Efficiency, Measurement, Ass↗
DOE OSTI2012

The SAS4A/SASSYS-1 Safety Analysis Code System

SAS4A/SASSYS-1 is a software simulation tool used to perform deterministic safety analysis of anticipated events as well as design basis and beyond design basis accidents for advanced nuclear reactors. This software can be used to assess the safety of a prescribed reactor design, but it cannot be used to configure a design to meet targeted performance objectives. Detailed, mechanistic models of steady-state and transient thermal,hydraulic, kinetic, and mechanical phenomena are employed to describe the response of the reactor core, the reactor primary and secondary coolant loops, the reactor control and protection systems and the balance-of-plant to accidents caused by changes in coolantflow, oss of heat rejection, or reactivity insertion. The consequences of single and double-fault accidents can be modeled, including fuel and coolant heating, fuel and cladding mechanical behavior, core reactivity feedbacks, coolant loop performance including natural circulation, and decay heat removal. Analyses are typically terminated upon demonstration of reactor and plant shutdown to permanently coolable conditions or upon violation of design basis margins. The objective of the analysis is to quantify accident consequences as measured by the transient behavior of system performance parameters, such as fuel and cladding temperatures, reactivity, and cladding strain. Originally developed for analysis of sodium cooled reactors with oxide fuel clad by stainless steel, the models In SAS4A/SASSYS-1 were subsequently extended and specialized to metallic fuel clad with advanced alloys and to several other coolant options including lead and LBE. METHOD OF SOLUTION: Each subassembly in the reactor core is represented with single or multiple-pin models. One channel represents one or more fuel pins and many channels are employed for a whole-core representation. Heat transfer in each pin is modeled with a two-dimensional (r/z) heat conduction equation. Single and two-phase coolant thermal-hydraulics are simulated with a unique, one-dimensional (axial) multiple-bubble liquid metal boiling model. The transient fuel and cladding mechanical oehavior models are integrated with fission product production,release, and transport models. UNIQUE FEATURES: The physical models in SAS4AJSASSYS-1 are highly detailed 1umerical representations of reactor accident conditions based on extensive laboratory and test reactor results. The models are specialized to liquid metal (sodium) cooled fast reactors with oxide or metallic fuel clad with stainless steel.

Source record↗
DOE OSTI2012

CBC Control Panel (Closed Brayton Cycle Control System) v 1.0

The CBC Control Panel is a LabviewTM Virtual Interface program that performs data acquisition, displays this data (over 100 channels), and provides control mechanisms for Closed Brayton Cycle (CBC) power conversion systems. The Program uses a state point control method that transitions the CBC from an IDLE/OFF state to RUNNING state, to SHUTDOWN, and ultimately bac~ to IDLE/OFF. During each state a set of rules control the behavior of the machine.

Source record↗
DOE OSTI2014

Determination of the Quantity of I-135 Released from the AGR Experiment Series

A series of three Advanced Gas Reactor (AGR) experiments have been conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). From 2006 through 2014, these experiments supported the development and qualification of the new U.S. tri structural isotropic (TRISO) particle fuel for Very High Temperature Reactors (VHTR). Each AGR experiment consisted of multiple fueled capsules, each plumbed for independent temperature control using a mix of helium and neon gases. The gas leaving a capsule was routed to individual Fission Product Monitor (FPM) detectors. For intact fuel particles, the TRISO particle coatings provide a substantial barrier to fission product release. However, particles with failed coatings, whether because of a minute percentage of initially defective particles, those which fail during irradiation, or those designed to fail (DTF) particles, can release fission products to the flowing gas stream. Because reactive fission product elements like iodine and cesium quickly deposit on cooler capsule components and piping structures as the effluent gas leaves the reactor core, only the noble fission gas isotopes of Kr and Xe tend to reach FPM detectors. The FPM system utilizes High Purity Germanium (HPGe) detectors coupled with a thallium activated sodium iodide NaI(Tl) scintillator. The germanium detector provides individual isotopic information, while the NaI(Tl) scintillator is used as a gross count rate meter. During irradiation, the 135mXe concentration reaching the FPM detectors is from both direct fission and by decay of the accumulated 135I. About ~2.5 hours after irradiation (ten 15.3 minute 135mXe half lives) the directly produced 135mXe has decayed and only the longer lived 135I remains as a source. Decay systematics dictate that 135mXe will be in secular equilibrium with its 135I parent, such that it’s production rate very nearly equals the decay rate of the parent, and its concentration in the flowing gas stream will appear to decay with the parent half life. This equilibrium condition enables the determination of the amount of 135I released from the fuel particles by measurement of the 135mXe at the FPM following reactor shutdown. In this paper, the 135I released will be reported and compared to similar releases for noble gases as well as the unexpected finding of 131I deposition from intentional impure gas injection into capsule 11 of experiment AGR 3/4.

AGR↗
DOE OSTI2016

Systems and methods for harvesting and storing materials produced in a nuclear reactor

Systems produce desired isotopes through irradiation in nuclear reactor instrumentation tubes and deposit the same in a robust facility for immediate shipping, handling, and/or consumption. Irradiation targets are inserted and removed through inaccessible areas without plant shutdown and placed in the harvesting facility, such as a plurality of sealable and shipping-safe casks and/or canisters. Systems may connect various structures in a sealed manner to avoid release of dangerous or unwanted matter throughout the nuclear plant, and/or systems may also automatically decontaminate materials to be released. Useable casks or canisters can include plural barriers for containment that are temporarily and selectively removable with specially-configured paths inserted therein. Penetrations in the facilities may limit waste or pneumatic gas escape and allow the same to be removed from the systems without over-pressurization or leakage. Methods include processing irradiation targets through such systems and securely delivering them in such harvesting facilities.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

Paddys Run Streambank Stabilization Project at the Fernald Preserve, Harrison, OH

The Fernald Preserve is a former uranium-processing plant that underwent extensive remediation pursuant to CERCLA and is now managed by the US DOE Office of Legacy Management. While remediation of buildings and soil contamination was completed in 2006, aquifer remediation is ongoing. Paddys Run is a second-order stream that runs to the south along the western side of the Fernald Preserve. The Paddys Run watershed encompasses nearly 41 km2 (16 mi2), including most of the Fernald site. Field personnel conducting routine site inspections in March 2014 observed that Paddys Run was migrating east via bank erosion into the “Pit 3 Swale,” an area of known surface-water contamination. The soil there was certified pursuant to site regulatory agreements and meets all final remediation levels. However, weekly surface-water monitoring is conducted from two puddles within the swale area, when water that exceeds the final remediation levels is present. Paddys Run had migrated east approximately 4 m (13 ft) in 2 years and was approximately 29 m (95 ft) from the sample location. This rapid migration threatened existing conditions that allowed for continued monitoring of the swale area and also threatened Paddys Run water quality. Therefore, DOE and regulators determined that the east bank of Paddys Run required stabilization. This was accomplished with a design that included the following components: relocation of approximately 145 m (475 ft) of streambed 9 m (30 ft) west, installation of a rock toe along the east bank, installation of two cross-vane in-stream grade-control structures, stabilization of a portion of the east bank using soil encapsulated lifts, and regrading, seeding, and planting within remaining disturbed areas. In an effort to take advantage of low-flow conditions in Paddys Run, construction was initiated in September 2014. Weather delays and subsurface flow within the Paddys Run streambed resulted in an interim shutdown of the project area in December 2014. Construction activities resumed in April 2015, with completion in November 2015. To date, this stabilization project has been successful. The regraded bank and streambed have remained stable, and no compromise to installed cross-vanes, the rock toe, or the soil encapsulated lifts has been observed.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

Examining Supply Chain Resilience for the Intermodal Shipment of Spent Nuclear Fuel and High Level Radioactive Materials

The U.S. Department of Energy (DOE) has a significant programmatic interest in the safe and secure routing and transportation of Spent Nuclear Fuel (SNF) and High Level Waste (HLW) in the United States, including shipments entering the country from locations outside U.S borders. In any shipment of SNF/HLW, there are multiple chains; a jurisdictional chain as the material moves between jurisdictions (state, federal, tribal, administrative), a physical supply chain (which mode), as well as a custody chain (which stakeholder is in charge/possession) of the materials being transported. Given these interconnected networks, there lies vulnerabilities, whether in lack of communication between interested stakeholders or physical vulnerabilities such as interdiction. By identifying key links and nodes as well as administrative weaknesses, decisions can be made to harden the physical network and improve communication between stakeholders. This paper examines the parallel chains of oversight and custody as well as the chain of stakeholder interests for the shipments of SNF/HLW and the potential impacts on systemic resiliency. Using the Crystal River shutdown location as well as a hypothetical international shipment brought into the United States, this paper illustrates the parallel chains and maps them out visually.

law↗
DOE OSTI2015

Recent QCD Results from the Tevatron

Four years after the shutdown of the Tevatron proton-antiproton collider, the two Tevatron experiments, CDF and DZero, continue producing important results that test the theory of the strong interaction, Quantum Chromodynamics (QCD). The experiments exploit the advantages of the data sample acquired during the Tevatron Run II, stemming from the unique pp initial state, the clean environment at the relatively low Tevatron instantaneous luminosities, and the good understanding of the data sample after many years of calibrations and optimizations. A summary of results using the full integrated luminosity is presented, focusing on measurements of prompt photon production, weak boson production associated with jets, and non-perturbative QCD processes.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2015

Plans for Deployment of Hollow Electron Lenses at the LHC for Enhanced Beam Collimation

Hollow electron lenses are considered as a possible means to improve the LHC beam collimation system, providing active control of halo diffusion rates and suppressing the population of transverse halos. After a very successful experience at the Tevatron, a conceptual design of a hollow e-lens optimized for the LHC was produced. Recent further studies have led to a mature preliminary technical design. In this paper, possible scenarios for the deployment of this technology at the LHC are elaborated in the context of the scheduled LHC long shutdowns until the full implementation of the HL-LHC upgrade in 2023. Possible setups of electron beam test stands at CERN and synergies with other relevant electron beam programmes are also discussed.

43 PARTICLE ACCELERATORS↗
DOE OSTI2015

The Machine Protection System for the Fermilab Accelerator Science and Technology Facility

The Machine Protection System (MPS) for the Fermilab Accelerator Science and Technology Facility (FAST) has been implemented and tested. The system receives signals from several subsystems and devices which conveys the relevant status needed to the safely operate the accelerator. Logic decisions are made based on these inputs and some predefined user settings which in turn controls the gate signal to the laser of the photo injector. The inputs of the system have a wide variety of signal types, encoding methods and urgencies for which the system is designed to accommodate. The MPS receives fast shutdown (FSD) signals generated by the beam loss system and inhibits the beam or reduces the beam intensity within a macropulse when the beam losses at several places along the accelerator beam line are higher than acceptable values. TTL or relay contact signals from the vacuum system, toroids, magnet systems etc., are chosen with polarities that ensure safe operation of the accelerator from unintended events such as cable disconnection in the harsh industrial environment of the experimental hall. A RS422 serial communication scheme is used to interface the operation permit generator module and a large number of movable devices each reporting multi-bit status. The system also supports operations at user defined lower beam levels for system conunissioning. The machine protection system is implemented with two commercially available off-the-shelf VMEbus based modules with on board FPGA devices. The system is monitored and controlled via the VMEbus by a single board CPU

Wu, Jinyuan↗
DOE OSTI2016

Parameterizing the Variability and Uncertainty of Wind and Solar in CEMs

We present current and improved methods for estimating the capacity value and curtailment impacts from variable generation (VG) in capacity expansion models (CEMs). The ideal calculation of these variability metrics is through an explicit co-optimized investment-dispatch model using multiple years of VG and load data. Because of data and computational limitations, existing CEMs typically approximate these metrics using a subset of all hours from a single year and/or using statistical methods, which often do not capture the tail-event impacts or the broader set of interactions between VG, storage, and conventional generators. In our proposed new methods, we use hourly generation and load values across all hours of the year to characterize the (1) contribution of VG to system capacity during high load hours, (2) the curtailment level of VG, and (3) the reduction in VG curtailment due to storage and shutdown of select thermal generators. Using CEM model outputs from a preceding model solve period, we apply these methods to exogenously calculate capacity value and curtailment metrics for the subsequent model solve period. Preliminary results suggest that these hourly methods offer improved capacity value and curtailment representations of VG in the CEM from existing approximation methods without additional computational burdens.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2014

Progress Towards Doubling the Beam Power at Fermilab's Accelerator Complex

After a 16 month shutdown to reconfigure the Fermilab Accelerators for high power operations, the Fermilab Accelerator Complex is again providing beams for numerous Physics Experiments. By using the Recycler to slip stack protons while the Main Injector is ramping, the beam power at 120 GeV can reach 700 KW, a factor of 2 increase. The progress towards doubling the Fermilab's Accelerator complex beam power will be presented.

43 PARTICLE ACCELERATORS↗
DOE OSTI2014

Impact of Distributed Wind on Bulk Power System Operations in ISO-NE

The work presented in this paper aims to study the impact of a range of penetration levels of distributed wind on the operation of the electric power system at the transmission level. This paper presents a case study on the power system in Independent System Operator New England. It is analyzed using PLEXOS, a commercial power system simulation tool. The results show that increasing the integration of distributed wind reduces total variable electricity generation costs, coal- and gas-fired electricity generation, electricity imports, and CO2 emissions, and increases wind curtailment. The variability and uncertainty of wind power also increases the start-up and shutdown costs and ramping of most conventional power plants.

17 WIND ENERGY↗