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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 1,099 records · Page 61
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Optimal SCR Control Using Data-Driven Models

We present an optimal control solution for the urea injection for a heavy-duty diesel (HDD) selective catalytic reduction (SCR). The approach taken here is useful beyond SCR and could be applied to any system where a control strategy is desired and input-output data is available. For example, the strategy could also be used for the diesel oxidation catalyst (DOC) system. In this paper, we identify and validate a one-step ahead Kalman state-space estimator for downstream NOx using the bench reactor data of an SCR core sample. The test data was acquired using a 2010 Cummins 6.7L ISB production engine with a 2010 Cummins production aftertreatment system. We used a surrogate HDD federal test procedure (FTP), developed at Michigan Technological University (MTU), which simulates the representative transients of the standard FTP cycle, but has less engine speed/load points. The identified state-space model is then used to develop a tunable cost function that simultaneously minimizes NOx emissions and urea usage. The cost function is quadratic and univariate, thus the minimum can be computed analytically. We show the performance of the closed-loop controller in using a reduced-order discrete SCR simulator developed at MTU. Our experiments with the surrogate HDD-FTP data show that the strategy developed in this paper can be used to identify performance bounds for urea dose controllers.

diesel, emmissions, control, environment, modeling↗
DOE OSTI2015

Summary of thermocouple performance during advanced gas reactor fuel irradiation experiments in the advanced test reactor and out-of-pile thermocouple testing in support of such experiments

High temperature gas reactor experiments create unique challenges for thermocouple-based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time-dependent change in composition and, as a consequence, a time-dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B) and tungsten-rhenium thermocouples (Type C). For lower temperature applications, previous experiences with Type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly, Type N thermocouples are expected to be only slightly affected by neutron fluence. Currently, the use of these nickel-based thermocouples is limited when the temperature exceeds 1000 deg. C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past 10 years, three long-term Advanced Gas Reactor experiments have been conducted with measured temperatures ranging from 700 deg. C - 1200 deg. C. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out-of-pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150 deg. C and 1200 deg. C for 2,000 hours at each temperature, followed by 200 hours at 1250 deg. C and 200 hours at 1300 deg. C. The standard Type N design utilizes high purity, crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including a Haynes 214 alloy sheath, spinel (MgAl{sub 2}O{sub 4}) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly thermocouple with hard-fired alumina insulation and a molybdenum sheath. The most current version of the High Temperature Irradiation Resistant Thermocouple, based on molybdenum/niobium alloys and developed at Idaho National Laboratory, was also tested. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Effects of ATR-2 Irradiation to High Fluence on Nine RPV Surveillance Materials

The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations that govern the operation of commercial nuclear power plants require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including pressurized thermal shock (PTS) in pressurized water reactors (PWR). In the irradiated condition, however, the fracture toughness of the RPV may be severely degraded, with the degree of toughness loss dependent on the radiation sensitivity of the materials. The available embrittlement predictive models and our present understanding of radiation damage are not fully quantitative, and do not treat all potentially significant variables and issues, particularly considering extension of operation to 80y.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Lessons Learned from Performing a Human Factors Engineering Validation of an Upgraded Digital Control System in a Nuclear Power Plant Control Room

The United States' fleet of nuclear power plants (NPPs) have exceeded their original 40-year licenses and many are seeking license renewals to continue operations into the future. To support continued operation the Department of Energy's (DOE) Light Water Reactor Sustainability Program (LWRS) provides support for utilities to maintain and upgrade their power production capabilities. Control room modernization is one focus area, which aims to modernization the analog control rooms with digital controls and human-system interfaces (HSIs). The Turbine Control System (TCS) is a non-safety related system undergoing modernization efforts at multiple plants and utilities. Considerable human factors research efforts, comprised of expert reviews and operator in the loop studies were performed to develop human factors compliant and usable TCS. This report documents the final phase of the human factors engineering program used to design and validate the new digital TCS. According to the NUREG-0711 review guide, one of the last steps of human factors engineering program prior to implementation is human factors validation. This report details the method, results and conclusions from the human factors validation study of a collaborating utilities TCS slated to be implemented in one of their plants. One control room team consisting of two control room supervisors and two reactor operators participated in the validation. All team members had participated in the BNP TCS training program foregoing the TCS validation. Performance measures included task performance, situation awareness and workload, expert observation of any negative influences of human-system interface features on control room team performance, and interviews with the participating operators. Performance criteria were based on the performance scores and the topics identified by interviews and expert observations. The results showed that the crew using the new TCS clearly could achieve their overall objectives in the test scenarios with sufficient situation awareness. The new TCS compared to the old system provided clear improvements of the control room work regarding shell and chest warming and regarding valve and other system testing. However, a number of Human Engineering Discrepancies (HEDs) were identified. HEDs were prioritized (level 1, 2 or 3) according to potential safety, plant availability, operability or personnel performance consequences. Since no priority 1 HEDs were identified there is no HFE reason not to continue implementation of the TCS. The TCS upgrade is acceptable if the Priority 2 HEDs are resolved and dispositioned in a satisfactory way. Priority 3 HEDs should be noted and considered for future modifications.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Progress Toward a Compact Fusion Reactor Using the Sheared-Flow-Stabilized Z-Pinch

The sheared-flow-stabilized (SFS) Z-pinch is a promising confinement concept for the development of a compact fusion reactor. The Z-pinch has been theoretically and experimentally shown to be stable to magnetohydrodynamic modes when sufficient radial shear of the axial flow is present. At the University of Washington, the Fusion Z-pinch Experiment (FuZE) research project examines scaling the SFS Z-pinch toward fusion conditions. The FuZE device produces long-duration, 50-cm-long pinches with measured ion and electron temperatures over 1 keV and number densities greater than 10 17 cm – 3 . Plasma properties are measured with a diagnostic suite that includes magnetic field probes, heterodyne quadrature interferometry, digital holographic interferometry, ion-Doppler spectroscopy, and fast framing photography. Neutrons are produced in the FuZE device when deuterium is injected along with the normal hydrogen or helium fueling species. Neutron generation is diagnosed using plastic scintillator detectors. The neutron production is sustained for 5 to 8 μs, thousands of times longer than the static Z-pinch instability growth time. We report measured neutron production is consistent with calculated theoretical values for thermonuclear yield at the observed plasma temperatures and scales with the square of the deuterium concentration. A preliminary reactor concept is designed to incorporate flowing liquid metal walls, which would serve as an electrode, a heat transfer fluid, a radiological shield, and a breeding blanket. Using a liquid metal wall could address several unresolved material and technology issues in existing fusion reactor designs.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Location-Dependent Mechanical Property Evaluation on Additively Manufacture Materials

This report summarizes research activities conducted at Argonne National Laboratory in support of the development, qualification and certification of additively manufactured (AM) metallic components to allow for innovative reactor design and licensing for the Transformational Challenge Reactor (TCR). Focus is on the evaluation of high temperature mechanical properties including creep and fatigue of 316L stainless steel manufactured by the laser powder bed fusion process. Creep behavior of AM 316L stainless steel was evaluated for rods printed by either or a combination of both lasers of the dual-laser system of a Concept Laser-M2 printer to examine the consistency across the build plate. Creep tests were conducted at temperatures of 550, 575, 600, and 650°C and stresses between 175 and 300 MPa using ASTM standard-sized specimens. The effect of heat treatments at different temperatures on creep properties of AM 316L SS was examined to understand the processing-microstructure-property relationship. Fatigue properties of AM 316L SS were investigated for two print geometries (rod and plate) and in two build orientations of printed plates. Locations of specimens in the build were carefully tracked such that the testing data can be directly related to location-specific in situ data to establish links between printing process, post-printing treatment, microstructure and mechanical properties. The work is to support the development of a digital platform informed approach to AM component qualification and certification for nuclear applications.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Argon Pumpout by ICRF Waves in C-Mod L- and I-mode Plasmas

Pumpout of argon ions by ICRF waves has been observed in C-Mod deuterium L- and I-mode plasmas that had a substantial hydrogen fraction. The effect is manifested by a reduction of core argon x-ray brightness up to a factor of 90% on time scales of tens of milliseconds following injection of ICRF power. For Ar^16+, the pumpout is strongest for hydrogen minority concentrations between 0.25 and 0.4, when the ICRF waves are not expected to result in minority heating. Modeling with the TORIC code suggests that the pumpout process occurs when the H/D mode conversion layer overlaps with the 2nd harmonic impurity resonance layer. The magnitude of the argon pumpout is independent of ICRF power above an apparent threshold of ~500 kW, independent of electron density and appears to decrease as the plasma current is increased. Potential application as a heavy impurity control tool in reactors is discussed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Development of Standardized Property Requirements, Measurement Methods, and Reporting Guidance for Coatings

The diverse research programs presently in progress around the world have advanced a large number of coated cladding concepts to various stages of maturity. Industry teams have developed leading coating compositions and geometries, but significant disparity exists in the importance placed on different coating properties, as well as the applicability and standardization of the various methods used to measure these properties. This milestone report provides an initial assessment of the most important coating properties and the experimental methods available for their determination to lay the foundation for standardized reporting and development of comprehensive datasets for modeling efforts. This report has determined that the residual stress of the coating is the most crucial property to investigate, yet it has received minimal attention to date due to a knowledge gap about the growth and formation of the brittle Laves phase under reactor conditions. The significance of residual stress of the coating is further enhanced by its connection and interdependence with other important properties like creep and coating failure mechanisms. This milestone proposes an importance rating for residual stress properties and their associated measurement techniques to fill in knowledge. Furthermore, techniques to investigate the outlined properties are discussed, and their capabilities and limitations are illustrated to provide guidance on their use.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Argon pumpout by ICRF waves in C-Mod L- and I-mode plasmas

Pumpout of argon ions by ICRF waves has been observed in C-Mod deuterium L- and I-mode plasmas that had a substantial hydrogen fraction. The effect is manifested by a reduction of core argon x-ray brightness up to a factor of 90% on time scales of tens of milliseconds following injection of ICRF power. For Ar 16+ , the pumpout is strongest for hydrogen minority concentrations between 0.25 and 0.4, when the ICRF waves are not expected to result in minority heating. Modeling with the TORIC code suggests that the pumpout process occurs when the H/D mode conversion layer overlaps with the 2nd harmonic impurity resonance layer. Here, the magnitude of the argon pumpout is independent of ICRF power above an apparent threshold of ~500 kW, independent of electron density and appears to decrease as the plasma current is increased. Potential application as a heavy impurity control tool in reactors is discussed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Cost- and Risk-Based Seismic Design Optimization of Nuclear Power Plant Safety Systems

Seismic analysis, design, and qualification of systems, structures, and components (SSCs) is a significant contributor to the capital cost of a nuclear power plant. To reduce capital costs of advanced nuclear power plants and make commercial nuclear energy more competitive, innovations are needed in their structural design and construction, and not just in the reactor core and associated systems. Seismic isolation has been identified as an important cost-cutting technology that enables standardization of equipment across various sites. This paper develops and demonstrates a cost- and risk-based seismic design optimization of a representative safety system in a nuclear power plant with the dual goals of minimizing overnight capital cost and meeting safety goals. The design optimization can also include component seismic isolation, in which case, the optimized design includes a set of equipment that needs to be seismically isolated to minimize capital cost. The open-source codes MASTODON and Dakota are used for seismic probabilistic risk assessment and design optimization, respectively. A generic nuclear facility with a safety system comprising SSCs that are common to nuclear power plants is considered for the demonstration of the design optimization and is assumed to be located at the Idaho National Laboratory site. Generic costs and seismic design cost functions are assumed for the SSCs of the safety system. The sum of the costs of the SSCs is minimized in the optimization process, while the risk of failure of the safety system is provided as a constraint. Furthermore, results show that the optimization process reduces capital costs significantly while automatically prioritizing the safety of SSCs that contribute most to the risk of the safety system.

42 ENGINEERING↗
DOE OSTI2016

Dakota Uncertainty Quantification Methods Applied to the CFD code Nek5000

Predictive modeling and simulation of the performance of a nuclear reactor and its fuel is a challenging task because of the large number of coupled physical phenomena that must be addressed. In addition to this intrinsic complexity, model uncertainty must be accounted for in any analysis if the model will be used to facilitate design or operational decisions which may impact safety and performance. Rigorous, structured uncertainty analyses are performed by first characterizing the model's input uncertainties and, by propagating the uncertainty through the model in order to estimate the output's uncertainty. This paper presents the state of advancement of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) project to characterize the uncertainty of the CFD code Nek5000 using the Dakota package. Nek5000 is a high order spectral element computational fluid dynamics (CFD) code developed at Argonne National Laboratory for high resolution spectral-filtered large eddy simulations (LESs) and unsteady Reynolds averaged Navier-Stokes (RANS) simulations. The Dakota package developed at Sandia National Laboratory can be integrated with many scientific and engineering codes to facilitate efficient, effective uncertainty quantification (UQ) and sensitivity analyses. Dakota is designed to interface with external packages/codes such as Nek5000 through a script-driven interface called fork. Such interface provides maximum flexibility to the users for pre- and post-processing tasks preceding and following each function evaluation. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Analysis of LOCA Scenarios in the NIST Research Reactor Before and After Fuel Conversion

An analysis has been done of hypothetical loss-of-coolant-accidents (LOCAs) in the research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The purpose of the analysis is to determine if the peak clad temperature remains below the Safety Limit, which is the blister temperature for the fuel. The configuration of the NBSR considered in the analysis is that projected for the future when changes will be made so that shutdown pumps do not operate when a LOCA signal is detected. The analysis was done for the present core with high-enriched uranium (HEU) fuel and with the proposed low-enriched uranium (LEU) fuel that would be used when the NBSR is converted from one to the other. The analysis consists of two parts. The first examines how the water would drain from the primary system following a break and the possibility for the loss of coolant from within the fuel element flow channels. This work is performed using the TRACE system thermal-hydraulic code. The second looks at the fuel clad temperature as a function of time given that the water may have drained from many of the flow channels and the water in the vessel is in a quasi-equilibrium state. The temperature behavior is investigated using the three-dimensional heat conduction code HEATING7.3. The results in all scenarios considered for both HEU and LEU fuel show that the peak clad temperature remains below the blister temperature.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Pre-conceptual design study of ASTRID core

In the framework of the ASTRID project at CEA, core design studies are performed at CEA with the AREVA and EDF support. At the stage of the project, pre-conceptual design studies are conducted in accordance with GEN IV reactors criteria, in particularly for safety improvements. An improved safety for a sodium cooled reactor requires revisiting many aspects of the design and is a rather lengthy process in current design approach. Two types of cores are under evaluation, one classical derived from the SFR V2B and one more challenging called CFV (low void effect core) with a large gain on the sodium void effect. The SFR V2b core have the following specifications: a very low burn-up reactivity swing (due to a small cycle reactivity loss) and a reduced sodium void effect with regard to past designs such as the EFR (around 2$ minus). Its performances are an average burn-up of 100 GWd/t, and an internal conversion ratio equal to one given a very good behavior of this core during a control rod withdrawal transient). The CFV with its specific design offers a negative sodium void worth while maintaining core performances. In accordance of ASTRID needs for demonstration those cores are 1500 MWth power (600 MWe). This paper will focus on the CFV pre-conceptual design of the core and S/A, and the performances in terms of safety will be evaluated on different transient scenario like ULOF, in order to assess its intrinsic behavior compared to a more classical design like V2B core. The gap in term of margin to a severe accident due to a loss of flow initiator underlines the potential capability of this type of core to enhance prevention of severe accident in accordance to safety demonstration. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Modernization of Technical Requirements for Licensing of Advanced Non-Light Water Reactors (Final Project Report)

Advanced (i.e., non-light-water) reactor technologies offer developers new opportunities to enhance the reliability, efficiency, and safety of nuclear power reactors through changes in fundamental design and operation. A variety of advanced design concepts are being developed that generally trend toward increased reliance on innovative, inherent, and passive safety features. Many of these features are substantially different from components now found in large light water reactors (LWRs) that dominate the commercial operating fleet. Similarly, non-light water reactor (non-LWR) suppliers are also pursuing market niches quite different from the regional baseload power generation sites typically associated with large LWR facilities.

42 ENGINEERING↗
DOE OSTI2016

Neutron characterization of UN/U-Si accident tolerant fuel prior to irradiation

Neutron imaging and tomographic reconstructions provide non-destructive three-dimensional characterization of nuclear fuel pellets that are typically about a cubic centimeter in volume with energy-resolved neutron imaging at a pulsed neutron source offering unique characterization not possible at a reactor source. Recent measurements at the Los Alamos Neutrons Science Center (LANSCE) studied a series of uranium nitride / uranium silicide ceramic composite pellets. This paper provides examples of spatially resolved measurements of cracks, isotopic enrichment and density with a (current) spatial resolution of approximately 75 microns. Phase and texture measurements were also performed but are not discussed in this paper. The UN/U-Si systems are under consideration as candidate systems for accident tolerant fuel. The examples are taken from measurements on a total of 9 pellets of differing UN/U-Si formulations (2 enrichment levels), U{sub 3}Si{sub 5} reference formulations (two enrichment levels) and pellets with known flaws. The {sup 235}U enrichments ranged from 0.2 to 8.8 wt.%. Pellets of the same composition are scheduled for irradiation in the Advanced Test Reactor this year. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Aleatoric and epistemic uncertainties in sampling based nuclear data uncertainty and sensitivity analyses

Sampling based uncertainty and sensitivity analyses due to epistemic input uncertainties, i.e. to an incomplete knowledge of uncertain input parameters, can be performed with arbitrary application programs to solve the physical problem under consideration. For the description of steady-state particle transport, direct simulations of the microscopic processes with Monte Carlo codes are often used. This introduces an additional source of uncertainty, the aleatoric sampling uncertainty, which is due to the randomness of the simulation process performed by sampling, and which adds to the total combined output sampling uncertainty. So far, this aleatoric part of uncertainty is minimized by running a sufficiently large number of Monte Carlo histories for each sample calculation, thus making its impact negligible as compared to the impact from sampling the epistemic uncertainties. Obviously, this process may cause high computational costs. The present paper shows that in many applications reliable epistemic uncertainty results can also be obtained with substantially lower computational effort by performing and analyzing two appropriately generated series of samples with much smaller number of Monte Carlo histories each. The method is applied along with the nuclear data uncertainty and sensitivity code package XSUSA in combination with the Monte Carlo transport code KENO-Va to various critical assemblies and a full scale reactor calculation. It is shown that the proposed method yields output uncertainties and sensitivities equivalent to the traditional approach, with a high reduction of computing time by factors of the magnitude of 100. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Application of MELCOR for Simulating Molten Salt Reactor Accident Source Terms

Molten Salt Reactor (MSR) systems can be divided into two basic categories: liquid-fueled MSRs in which the fuel is dissolved in the salt, and solid-fueled systems such as the Fluoride-salt-cooled High-temperature Reactor (FHR). The molten salt provides an impediment to fission product release as actinides and many fission products are soluble in molten salt. Nonetheless, under accident conditions, some radionuclides may escape the salt by vaporization and aerosol formation, which may lead to release into the environment. Here, we present recent enhancements to MELCOR to represent the transport of radionuclides in the salt and releases from the salt. Some soluble but volatile radionuclides may vaporize and subsequently condense to aerosol. Insoluble fission products can deposit on structures. Thermochimica, an open-source Gibbs Energy Minimization (GEM) code, has been integrated into MELCOR. With the appropriate thermochemical database, Thermochimica provides the solubility and vapor pressure of species as a function of temperature, pressure, and composition, which are needed to characterize the vaporization rate and the state of the salt with fission products. Since thermochemical databases are still under active development for molten salt systems, thermodynamic data for fission product solubility and vapor pressure may be user specified. This enables preliminary assessments of fission product transport in molten salt systems. In this paper, we discuss modeling of soluble and insoluble fission product releases in a MSR with Thermochimica incorporated into MELCOR. Separate-effects experiments performed as part of the Molten Salt Reactor Experiment in which radioactive aerosol was released are discussed as needed for determining the source term.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

RELAP5-3D Resolution of Known Restart/Backup Issues

The state-of-the-art nuclear reactor system safety analysis computer program developed at the Idaho National Laboratory (INL), RELAP5-3D, continues to adapt to changes in computer hardware and software and to develop to meet the ever-expanding needs of the nuclear industry. To continue at the forefront, code testing must evolve with both code and industry developments, and it must work correctly. To best ensure this, the processes of Software Verification and Validation (V&V) are applied. Verification compares coding against its documented algorithms and equations and compares its calculations against analytical solutions and the method of manufactured solutions. A form of this, sequential verification, checks code specifications against coding only when originally written then applies regression testing which compares code calculations between consecutive updates or versions on a set of test cases to check that the performance does not change. A sequential verification testing system was specially constructed for RELAP5-3D to both detect errors with extreme accuracy and cover all nuclear-plant-relevant code features. Detection is provided through a “verification file” that records double precision sums of key variables. Coverage is provided by a test suite of input decks that exercise code features and capabilities necessary to model a nuclear power plant. A matrix of test features and short-running cases that exercise them is presented. This testing system is used to test base cases (called null testing) as well as restart and backup cases. It can test RELAP5-3D performance in both standalone and coupled (through PVM to other codes) runs. Application of verification testing revealed numerous restart and backup issues in both standalone and couple modes. This document reports the resolution of these issues.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗