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At least 1,675 records · Page 93
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Optimal Control of Biomass Feedstock Processing System Under Uncertainty in Biomass Quality

Planning of biorefinery operations is complicated by the stochastic nature of physical and chemical characteristics of biomass feedstock, such as, moisture level and carbohydrate content. Biomass characteristics affect the performance of the equipment which feed the reactor and the efficiency of the conversion process in a biorefinery. We propose a stochastic optimization model to identify a blend of feedstocks, inventory levels, and operating conditions of equipment to ensure a continuous flowing of biomass to the reactor while meeting the requirements of the biochemical conversion process. We propose a sample average approximation (SAA) of the model, and develop an efficient algorithm to solve the SAA model. A feedstock preprocessing process consists of two-stage grinding and pelleting is used to develop a case study. Extensive numerical analysis are conducted which lead to a number of observations. Our main observation is that sequencing bales based on moisture level and carbohydrate content leads to robust solutions that improve processing time and processing rate of the reactor. We provide a number of managerial insights that facilitate the implementation of the model proposed. Note to Practitioners—This paper is motivated by the challenges faced in the bioenergy industry. The focus of this paper is on plants which use the biochemical conversion process to generate liquid fuels. It has been observed that variations in biomass characteristics, such as moisture content, cause variations in feeding of the system which lead to under-utilization of equipment. A requirement of biochemical conversion process is to maintain the carbohydrate content of biomass processed by the reactor, larger than a threshold. We propose a model that identifies the inventory levels and operating conditions of equipment to ensure a continuous flowing of biomass to the reactor. The goal is to improve equipment utilization while satisfying the requirements of the conversion process. The model is tested using real-life data. We found out that by sequencing bales based on moisture level and carbohydrate content, a plant can reduce variability in the system leading to improved system reliability, higher processing rates of the reactor, and higher throughput.

09 BIOMASS FUELS↗
DOE OSTI2026

Programmed synthesis of mesoporous protein crystals in cellular reactors

Protein crystals are naturally derived mesoporous materials with versatile structures and physicochemical properties. Here we introduce an intracellular synthesis platform that enables controllable and programmable protein crystallization. In live cells, we show that, after initial nucleation, steady protein expression governs crystal growth, yielding predictable, tunable dynamics in live cells. Exploiting this feature, we combined HaloTag and click chemistries to achieve modular, programmable immobilization of diverse guest materials with spatial patterning down to ~100 nm resolution. We further demonstrated the sequential release of immobilized materials in physiologically relevant fluids. As a proof of concept, we programmed particles to carry human fibroblast growth factors in distinct layers, which elicited designed oscillatory Akt signalling patterns in cell culture. Finally, this work outlines a programmable method for producing mesoporous materials, with possible applications in catalysis and biomedicine.

Yang, Hongru [Johns Hopkins Univ., Baltimore, MD (↗
DOE OSTI2016

Shift Verification and Validation

This documentation outlines the verification and validation of Shift for the Consortium for Advanced Simulation of Light Water Reactors (CASL). Five main types of problems were used for validation: small criticality benchmark problems; full-core reactor benchmarks for light water reactors; fixed-source coupled neutron-photon dosimetry benchmarks; depletion/burnup benchmarks; and full-core reactor performance benchmarks. We compared Shift results to measured data and other simulated Monte Carlo radiation transport code results, and found very good agreement in a variety of comparison measures. These include prediction of critical eigenvalue, radial and axial pin power distributions, rod worth, leakage spectra, and nuclide inventories over a burn cycle. Based on this validation of Shift, we are confident in Shift to provide reference results for CASL benchmarking.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

A Spitzer Model-C Stellarator fusion reactor

Tokamaks are affected by two serious problems. The first problem is the radiation damage to the internal reactor structures by energetic neutrons carrying 80% of the total fusion energy that are generated by the deuterium–tritium fusion cycle that tokamaks are planning to use. The second problem is a loss of plasma control because of the onset of instabilities, causing the sudden shutdown of the tokamak itself. Certainly, no electric utility knowing about these disruptive events would dare to use a tokamak reactor for powering the turbines of its power plant! This paper describes a new fusion reactor based on the Model-C Stellarator that would not be affected by these fatal phenomena.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Impact of control blade insertion on the deformation behavior of SiC-SiC channel boxes in BWRs

This article describes the analysis of distortion of a silicon carbide fiber-reinforced, silicon carbide matrix (SiC-SiC) composite channel box under in-reactor conditions of a boiling water reactor (BWR). The BWR core has significant gradients in the fast neutron flux across the channel box due to the presence of water rods within the fuel assemblies, and these gradients increase further with the insertion of control blades. As a result of the temperature and neutron flux dependent irradiation-induced swelling of SiC, the SiC-SiC composite channel box can undergo distortion. In this work, we evaluate the SiC-SiC channel box distortion for three different control blade positions. This analysis is based on the neutron flux and temperature distributions in the BWR core calculated using the neutronics code MPACT and thermal-hydraulics code CTF. This calculation is coupled through temperature feedback. Subsequently, we have performed structural analysis based on the calculated neutron flux and temperature distributions to determine the deformation and stress development in the channel box. The structural analysis was performed using the fuel performance modeling code BISON and the commercial finite element analysis software Abaqus. The results indicate that large gradients in fast neutron flux (up to 35–40% across a single axial level) will develop across the channel box. Due to these gradients, the channel box will undergo time-dependent bending for all the control blade positions in the assembly. The time-dependent bowing behavior is dominated by the transient swelling of SiC-SiC material under non-uniform neutron flux, and changes with variation in the control blade position. The bending will cause temporary interference between the channel box and control blade, and the interference is expected to be most severe for the fully inserted control blade position. The developed stresses due to differential swelling in the channel box exceed the proportional limit stress of the material, which may cause matrix microcracking in the channel box. However, the stresses remain below the tensile strength of the material, and therefore, development of a full, through-thickness crack in the channel box is not expected. Finally, further work is recommended to explore and evaluate the mitigation strategies.

42 ENGINEERING↗
DOE OSTI2020

Cylindricity Sensitivity Thermal Model of the AGR-5/6/7 Experiment in the Advanced Test Reactor

The AGR-5/6/7 experiment is currently being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory and is approximately 70% complete. Several fuel and material irradiation experiments have been planned for the U.S. Department of Energy Advanced Gas Reactor Fuel Development and Qualification Program, which supports the development and qualification of tristructural isotropic (TRISO) coated particle fuel for use in high-temperature gas-cooled reactors. The goals of these experiments are to provide irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, support development of fuel performance models and codes, and provide irradiated fuel and materials for post-irradiation examination and safety testing. Originally planned and named as separate fuel experiments, but subsequently combined into a single test train, AGR-5/6/7 is testing low-enriched uranium oxycarbide TRISO fuel. The AGR-5/6/7 test train has 5 capsules with thermocouples and independent gas control mixtures. Unique to this paper is a sensitivity study concerning the cylindricity of the graphite holders containing the fuel compacts and their eccentricity in relation to the stainless-steel capsule walls. Each capsule has small nubs on the outside used for centering the graphite holder inside the stainless-steel capsule with a small gas gap used to control temperature. Due to machining tolerances of these nubs and vibration wearing the nubs down when the experiment is running in the reactor, the possibility exists that the holder may move around radially. Each capsule is equipped with several thermocouples placed at various radii and depths within each graphite holder. This paper will show the sensitivity of offsetting the graphite holder for various radii in 45-degree increments around the circle with the objective of minimizing the difference between the measured thermocouples and the modeled thermocouple temperatures. Separate gas mixtures of helium/neon are introduced into this gas gap between the holder and capsule wall and changed as necessary to maintain the desired thermocouple temperatures to keep the fuel compacts at constant temperature as the nuclear reactor conditions change. The goal of the sensitivity study is to find a radius and an angle to offset the holder from perfectly centered for each of the five capsules separately. The complex thermal model includes fission heating, gamma heating, radiation heat transfer, and heat transfer via conduction and radiation across the control gaps. Subroutines linked to the thermal model offer an easy method to offset the graphite holder from the capsule walls without remeshing the entire model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

CS-Studio Scan System Parallelization

For several years, the Control System Studio (CS-Studio) Scan System has successfully automated the operation of beam lines at the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) and Spallation Neutron Source (SNS). As it is applied to additional beam lines, we need to support simultaneous adjustments of temperatures or motor positions. While this can be implemented via virtual motors or similar logic inside the Experimental Physics and Industrial Control System (EPICS) Input/Output Controllers (IOCs), doing so requires a priori knowledge of experimenters requirements. By adding support for the parallel control of multiple process variables (PVs) to the Scan System, we can better support ad hoc automation of experiments that benefit from such simultaneous PV adjustments.

43 PARTICLE ACCELERATORS↗
DOE OSTI2024

Role of turbulent separatrix tangle in the improvement of the integrated pedestal and heat exhaust issue for stationary-operation tokamak fusion reactors

Abstract The magnetic separatrix surface is designed to provide the final and critical confinement to the hot stationary-operation core plasma in modern tokamak reactors in the absence of an external magnetic perturbation (MP) or transient magneto-hydrodynamic perturbation, while diverting the exhaust heat to divertor plates. All the stationary operational boundary plasma studies and reactor designs have been performed under this assumption. However, there has been a long-standing suspicion that a stationary-operation tokamak plasma even without external MPs or edge localized modes (ELMs) activities may not have a stable closed separatrix surface, especially near the magnetic X-point. Here, the first gyrokinetic numerical observation is reported that the divertor separatrix surface, due to homoclinic tangles caused by intrinsic electromagnetic turbulence, is not a stable closed surface in a stationary operation phase even without MPs or ELMs. Unlike the MP- or ELM-driven homoclinic tangles that could cause deleterious effects to core confinement or divertor plates, it is found that the micro-turbulence driven homoclinic tangles could connect the divertor plasma to the pedestal plasma in a constructive way by broadening the divertor heat-exhaust footprint and weakening the pedestal slope to the ELM-safe direction. Micro-turbulent homoclinic tangles can open a new research direction in understanding and controlling these two most troublesome and non-locally connected edge-plasma issues in a tokamak fusion reactor.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Wide-range structurally optimized channel for monitoring the certified power of small-core reactors

The results of tests of a prototype version of a channel for monitoring the certified power of small-core reactors performed at the BR-K1 reactor at the All-Russian Scientific Research Institute of Experimental Physics are reported. An SNM-11 counter and commercial KNK-4 and KNK-3 compensated ion chambers were used as neutron detectors in the tested channel, and certified NCMM and CCMM measurement modules controlled by a PC with specialized software were used as measuring instruments. The specifics of metrological assurance of calibration of the channel in the framework of reactor power monitoring are discussed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

In-field Calibration of a Fast Neutron Collar for the Measurement of Fresh PWR Fuel Assemblies

A new neutron collar has been designed for the measurement of fresh LEU fuel assemblies. This collar uses “fast mode” measurement to reduce the effect of burnable poison rods on the assay and thus reduce the dependence on the operator’s declaration. The new collar design reduces effect of poison rods considerably. Instead of 12 pins of 5.2% Gd causing a 20.4% effect, as in the standard thermal mode collar, they only cause a 3.2% effect in the new collar. However it has higher efficiency so that reasonably precise measurements can be made in 25 minutes, rather than the 1 hour of previous collars. The new collar is fully compatible with the use of the standard data collection and analysis code INCC. This report describes the calibration that was made with a mock-up assembly at Los Alamos National Laboratory and with actual assemblies at the AREVA Fuel fabrication Plant in Lingen, Germany.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI1996

Nuclear Safety [Vol. 37, No. 3, July-September 1996]

Nuclear Safety is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS 181 Attitudes and Practices Regarding Disposal of Liquid Nuclear Waste at Clinton Laboratories in the Very Early Years: A Historical Analysis, S. H. Stow; 202 Off-Site Nuclear Emergency Management in Germany Under the Auspices of the Federal Structure, A. Bayer, S. Bittner, and H. Korn; DESIGN FEATURES: 211 Twenty-Fourth DOE/NRC Nuclear Air-Cleaning and Treatment Conference, R. R. Bellamy, J. J. Hayes, R. D. Porco, and M. W. First; OPERATING EXPERIENCES: 222 Explosion in the Tomsk-7 Reprocessing Plant on April 6, 1993, M. L. Hyder, W. G. Lussie, and F. E. Witmer; 234 Reactor Shutdown Experience, Compiled by J. W. Cletcher; U.S. NUCLEAR REGULATORY COMMISSION INFORMATION AND ANALYSES: 237 Assessment of Spent Fuel Cooling, J. G. Ibarra, W. R. Jones, G. F. Lanik, H. L. Ornstein, and S. V. Pullani; RECENT DEVELOPMENTS: 256 Reports, Standards, and Safety Guides, D. S. Queener; 260 Proposed Rule Changes as of June 30, 1996; ANNOUNCEMENTS: 210 Tennessee Industries Week; 221 21st International Symposium on The Scientific Basis for Nuclear Waste Management; 259 IAEA Programme of Scientific Meetings for 1997; 270 PSAM 4 (International Conference on Probabilistic Safety Assessment and Management); 265 The Authors; 268 Errata.

05 NUCLEAR FUELS↗
DOE OSTI2022

Considerations of Radiation-Hardened Electronics for Alpha Detectors in Molten Salt Reactors

Along with advanced generations of reactors quickly becoming reality, the need for instrumentation to provide reliable information for their safety and operation is also becoming relevant. Instrumentation-wise, molten salt reactors (MSRs) represent an especially challenging case in which detectors must survive extreme temperatures, high amounts of radiation, and a highly corrosive coolant. Adding to the complexity, tight material control and accounting must be maintained with any special nuclear material (SNM), a practice made more difficult by the nature of a loop of flowing fuel. Alpha spectroscopy offers a unique window into investigating the isotopic concentrations of actinides within just such a molten salt environment. The characteristic α emissions of specific actinide isotopes in the salt enable quantification of their concentrations. Semiconductor devices based on wide bandgap materials such as silicon carbide and gallium oxide can withstand the extreme temperatures within MSRs and offer good chemical stability with the corrosive salt. Collaborators at The Ohio State University are fabricating and testing a β gallium oxide α particle sensor that could identify actinide concentrations within the fuel salt to help improve the material control and accounting required for complicated system inherent to MSRs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Modeling and analysis of an AMB system with flexible rotor in HTR-10GT - 18550

Active magnetic bearing (AMB) has been widely used in high-temperature gas-cooled reactor (HTR) for its specific properties: free of contact, absence of lubrication, capability of rotating at high speed. AMB system is inherently unstable since positive feedback exists in the open loop system and, thus, requires feedback control. From the control view of point, accurate system model is essential for designing of an effective controller especially when the flexible gyroscopic rotor is involved. In the paper, a flexible rotor and an open-loop AMB system model were established for the purpose of controller design. Dynamic characteristics of the undamped flexible rotor are analyzed including modal vibration analysis and undamped natural frequencies. Additionally, some characteristics of open-loop AMB system are investigated such as relationship between support stiffness and natural frequencies, and the system frequency response. The analysis results in this paper will provide valuable information for AMB system feedback controller design. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1984

Nuclear-electric power in space

Prospective missions requiring large power supplies that might be satisfied with space nuclear reactors (SNR) are discussed, along with design concepts and problems and other potential high-power space systems. Having a minimum economic output of 10 kWe, SNR seem well-suited as the power sources for DBS systems, space-based ATC systems manned planetary missions, an expanding Space Station, materials processing, and outer planets missions. SNR avoid the large area problems of solar cell arrays, short lifetimes of thermionic converters, and vibration and heat control in Stirling engines. Design problems exist for SNR in the heat transfer and rejection systems, radioactive emissions and degradation of reactor materials, and size. The latter is a function of Shuttle payload constaints and raises the possibility of having to load the fuel while in orbit. The earliest operational date of SNRs is projected for the early 1990s, if progress is good in the current SP-100 program.

Truscello, V. C.↗
DOE OSTI2026

A single-zone zero-dimensional study of HCCI combustion of methanol dehydration products to enable ignition of direct-injected methanol

Methanol is an alternative fuel gaining traction in the maritime sector. Its direct adoption, however, is accompanied by a unique set of technical challenges, such as low cetane number and high latent heat of vaporization. An approach to overcome these challenges is being developed at the US Department of Energy’s Oak Ridge National Laboratory, where onboard generation of dimethyl ether (DME) via catalytic dehydration of methanol can be used to assist in the mixing controlled combustion of direct-injected (DI) methanol. The generated mixture from this dehydration process can be premixed with intake air to condition the cylinder via. homogeneous charge compression ignition (HCCI) for subsequent DI methanol. In this preliminary work, various catalyst or reactor conversion efficiencies were simulated (using bottles) at constant DME and water flow at low load on a single-cylinder marine-variant of a CAT® C18 18 L engine with a 145 mm bore. To substantiate the experimental findings, a zero-dimensional engine model was developed in Cantera using a DME mechanism with 79 species and 658 reactions. Results presented include experimental and simulation heat release rate comparisons, species evolution information, and constant volume ignition delay (ID) for DI methanol with and without background species from HCCI of the premixed products from different reactor efficiencies. The results suggest that thermal effects dominate the DI methanol ignition process, and this work provides a chemical kinetic foundation or guideline for developing future control schemes.

Tyrewala, Daanish [ORNL] (ORCID:0000000208599324)↗
DOE OSTI2022

Evaluation of Joint Cyber/Safety Risk in Nuclear Power Systems

This report presents an analysis of the Emergency Core Cooling System (ECCS) for a generic Boiling Water Reactor (BWR)-4 NPP. The Electric Power Research Institute (EPRI) developed Hazards and Consequences Analysis for Digital Systems (HAZCADS) process is applied to the ECCS and its subsystems to identify unsafe control actions (UCAs) which act as possible cyber events of concern. The analysis is performed for two design basis events: Small-break Loss of Coolant Accident (SLOCA) and general transients (TRANS), such as unintended reactor trip. In previous work, HAZCADS UCAs were combined with other cyber-attack analysis to develop a risk-informed approach; however, this was for a single system. This report explores advanced systems engineering modeling approaches to model the interactions between digital assets across multiple systems which may be targeted by cyber adversaries. The complex and interdependent design of digital systems has the potential to introduce emergent cyber properties that are generally not covered by hazard analyses nor formal nuclear Probabilistic Risk Assessment (PRA). The R&D and supporting analysis presented here explores approaches to predict and manage how interdependent system properties effect risk. To show the potential impact of a successful cyber-attack to formal PRA event tree probabilities, HAZCADS analysis was also used. HAZCADS was also used to model the automatic depressurization system (ADS) automatic actuation. This analysis extended to an integrated system analysis for common-cause failure (CCF). In this aspect, the HAZCADS analysis continued by analyzing plant design details for system connectivity in support of critical plant functions. A dependency matrix was developed to depict the integrated functionality of the interconnected systems. Areas of potential CCF are indicated. Future work could include adversary attack development to show how CCF could be caused, resulting in PRA events. Across the multiple systems that comprise the ECCS, the analysis shows that the change in such probabilities was very different between systems. This indicates that some systems have a larger potential risk impact from successful cyber-attack or digital failure, which indicates a need for these systems to have a higher priority for design and defensive measures. Furthermore, we were able to establish that a risk analysis using any arbitrary threat model establishes an ordering of components with regard to cyber-risk. This ordering can be used to influence the overall system design with an eye to lowering risk, or as a way to understand real-time risk to operational systems based on a current threat landscape. Expert knowledge of both the analysis process and the system being analyzed is required to perform a HAZCADS analysis. The need for a tiered risk analysis is demonstrated by the results of this report.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Creep-Fatigue Life Prediction of Alloy 617 at 950 deg. C using an Effective Hold Time Approach - 18511

Time Fraction Rule for calculating creep damage fraction is suitable for predicting creep-fatigue life of Alloy 617 at 950 deg. C. However, purely strain-controlled loading waveforms cannot produce creep-dominant interaction due to rapid stress-relaxation of Alloy 617. This study employs a novel loading profile with a combination of strain and force-control to produce varying proportions of creep and fatigue damage. The Time Fraction Rule is found to be inaccurate for life prediction for this novel loading waveform. A new method for calculating the creep damage fraction is proposed with the objective of predicting creep-fatigue life for traditional purely strain-controlled and novel hybrid control loading waveforms. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗