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At least 865 records · Page 48
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Method and apparatus for virtualizing the micro-op cache

Systems, apparatuses, and methods for virtualizing a micro-operation cache are disclosed. A processor includes at least a micro-operation cache, a conventional cache subsystem, a decode unit, and control logic. The decode unit decodes instructions into micro-operations which are then stored in the micro-operation cache. The micro-operation cache has limited capacity for storing micro-operations. When new micro-operations are decoded from pending instructions, existing micro-operations are evicted from the micro-operation cache to make room for the new micro-operations. Rather than being discarded, micro-operations evicted from the micro-operation cache are stored in the conventional cache subsystem. This prevents the original instruction from having to be decoded again on subsequent executions. When the control logic determines that micro-operations for one or more fetched instructions are stored in either the micro-operation cache or the conventional cache subsystem, the control logic causes the decode unit to transition to a reduced-power state.

Kalamatianos, John↗
DOE OSTI2023

Filtering micro-operations for a micro-operation cache in a processor

A processor includes a micro-operation cache having a plurality of micro-operation cache entries for storing micro-operations decoded from instruction groups and a micro-operation filter having a plurality of micro-operation filter table entries for storing identifiers of instruction groups for which the micro-operations are predicted dead on fill if stored in the micro-operation cache. The micro-operation filter receives an identifier for an instruction group. The micro-operation filter then prevents a copy of the micro-operations from the first instruction group from being stored in the micro-operation cache when a micro-operation filter table entry includes an identifier that matches the first identifier.

Scrbak, Marko↗
DOE OSTI2012

Fuel Pond Sludge - Lessons Learned from Initial De-sludging of Sellafield's Pile Fuel Storage Pond - 12066

The Pile Fuel Storage Pond (PFSP) at Sellafield was built and commissioned between the late 1940's and early 1950's as a storage and cooling facility for irradiated fuel and isotopes from the two Windscale Pile reactors. The pond was linked via submerged water ducts to each reactor, where fuel and isotopes were discharged into skips for transfer along the duct to the pond. In the pond the fuel was cooled then de-canned underwater prior to export for reprocessing. The plant operated successfully until it was taken out of operation in 1962 when the First Magnox Fuel Storage Pond took over fuel storage and de-canning operations on the site. The pond was then used for storage of miscellaneous Intermediate Level Waste (ILW) and fuel from the UK's Nuclear Programme for which no defined disposal route was available. By the mid 1970's the import of waste ceased and the plant, with its inventory, was placed into a passive care and maintenance regime. By the mid 1990s, driven by the age of the facility and concern over the potential challenge to dispose of the various wastes and fuels being stored, the plant operator initiated a programme of work to remediate the facility. This programme is split into a number of key phases targeted at sustained reduction in the hazard associated with the pond, these include: - Pond Preparation: Before any remediation work could start the condition of the pond had to be transformed from a passive store to a plant capable of complex retrieval operations. This work included plant and equipment upgrades, removal of redundant structures and the provision of a effluent treatment plant for removing particulate and dissolved activity from the pond water. - Canned Fuel Retrieval: Removal of canned fuel, including oxide and carbide fuels, is the highest priority within the programme. Handling and export equipment required to remove the canned fuel from the pond has been provided and treatment routes developed utilising existing site facilities to allow the fuel to be reprocessed or conditioned for long term storage. - Sludge Retrieval: In excess of 300 m{sup 3} of sludge has accumulated in the pond over many years and is made up of debris arising from fuel and metallic corrosion, wind blown debris and bio-organic materials. The Sludge Retrieval Project has provided the equipment necessary to retrieve the sludge, including skip washer and tipper machines for clearing sludge from the pond skips, equipment for clearing sludge from the pond floor and bays, along with an 'in pond' corral for interim storage of retrieved sludge. Two further projects are providing new plant processing routes, which will initially store and eventually passivate the sludge. - Metal Fuel Retrieval: Metal Fuel from early Windscale Pile operations and various other sources is stored within the pond; the fuel varies considerably in both form and condition. A retrieval project is planned which will provide fuel handling, conditioning, sentencing and export equipment required to remove the metal fuel from the pond for export to on site facilities for interim storage and disposal. - Solid Waste Retrieval: A final retrieval project will provide methods for handling, retrieval, packaging and export of the remaining solid Intermediate Level Waste within the pond. This includes residual metal fuel pieces, fuel cladding (Magnox, aluminium and zircaloy), isotope cartridges, reactor furniture, and miscellaneous activated and contaminated items. Each of the waste streams requires conditioning to allow it to be and disposed of via one of the site treatment plants. - Pond Dewatering and Dismantling: Delivery of the above projects will allow operations to progressively remove the radiological inventory, thereby reducing the hazard/risk posed by the plant. This will then allow subsequent dewatering of the pond and dismantling of the structure. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Radionuclide Inventory Calculations for the Calcined Solid Storage Facility Performance Assessment - 19292

Originally named the Idaho Chemical Processing Plant when it was constructed in the early 1950's, the Idaho Nuclear Technology and Engineering Center reprocessed SNF to recover highly enriched uranium and other nuclear-related products for the U.S. DOE. Reprocessing SNF generated millions of gallons of aqueous radioactive waste. Calcination was a process developed at the INL Site to convert liquid waste into a solid, granular form from 1963 to 2000. The calcination process produced a safer product for storage while reducing the volume of stored liquid waste by an average factor of seven. Calcine is currently stored in the Calcined Solids Storage Facility (CSSF). The CSSFs are often called 'bin sets' because each CSSF (CSSFs 1 through 6) contains several stainless-steel storage bins housed in a concrete vault. The CSSF is an interim waste storage facility. In the future, calcine will be retrieved, treated (if necessary), and sent to an off-Site facility for storage or disposal. To close the CSSF, the bulk of the calcine will be removed and stored or disposed of elsewhere. The bins and vaults will then be placed in a configuration that is protective of human health and the environment. For purposes of the performance assessment, the depth of the residual calcine at closure was assumed to be 5-cm in each bin. The 5-cm of residual calcine in each bin is assumed to be mixed in grout during closure. The activities of the residual radionuclides are the inventory used for the dose assessment calculations for the CSSF performance assessment. The objective of this analysis was to calculate the radionuclide activity in the residual calcine remaining in the CSSF after closure. The purpose of this paper is to present the approach used to calculate the radionuclide inventory of the calcine currently stored in CSSF and the residual calcine in the CSSF after closure using data from the calcined waste Historical Processing Model (HPM). Limited sampling and analysis of the calcine in the CSSF bins was conducted in the past. Future similar calcine sampling and analysis efforts would be costly, generate additional mixed waste, and may not provide the desired detailed calcine composition data. However, during the later years of calcining, an alternative to calcine sampling and analysis was developed and the information was captured in the HPM. The HPM uses multiple sources to provide definitive calcine volume, mass, and composition, including: compositions and volumes of the liquid wastes that were calcined, ORIGEN2 - and ORIGEN-ARP-based radioactivity calculations, calciner and CSSF operating data, and process knowledge of INTEC fuel and waste-processing systems. HPM calcine composition was used to develop an inventory of radionuclides for use in the performance assessment evaluation for closure of CSSF. There are 43 bins that contain calcine, with six to twelve segments in each bin for a total of 337 bin segments for which the HPM calculates the average calcine composition. The HPM calculates the composition of each of the nearly 7,000 batches of material (startup bed, nonradioactive material, and radioactive waste) charged to the calciners. The calculations use the CSSF dimensions, including the CSSF number of bins, heights, outside diameters, annular air space diameter (if necessary), and volume of calcine currently in the bins. For each of the CSSFs, for a 5-cm residual, the percent of total calcine was calculated along with the residual activity for the radionuclides. Residual activity was estimated for 148 different radionuclides. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Cradle-to-Grave: Assessment Methods for Hazardous Waste Management

The Resource Conservation and Recovery Act (RCRA) was passed to set up a framework for the proper management of waste. The 40CFR260-273 regulation established the regulatory framework to handle solid and hazardous waste, from the point of generation to its disposal; or cradle-to-grave management. Per these regulations, a hazardous waste generator is responsible for its waste from initial generation through its ultimate disposal and beyond. Nothing will relinquish a generator of this responsibility. There is no expiration date or time limit, and hiring a someone else to transport and dispose of waste does not transfer responsibility. If an incident occurs at any point during the life cycle of a waste, the generator can be held liable. The first step to compliance with RCRA is proper characterization of the waste. Waste may either be listed specifically or may show a characteristic as defined in the regulation: Is my waste hazardous? Why is it Hazardous? What should I do as a generator? Plan chemical use: - Create a chemical inventory for your research project. - Use existing chemicals if available instead of ordering new ones. - Brainstorm methods that will help to reduce toxicity and minimize waste. Waste Inventory: - Label all waste bottles with: ID no., What is it, Whose is it, Hazards, and Date. - Add waste containers into the inventory log. - Initiate/complete the container sheet to describe the waste. Satellite Areas: - Store waste in containers compatible with the waste with lids on tight. - Store waste containers properly and segregate them according to their specific hazards. - Use secondary containment for liquids. Store acids and bases separately; apart from cyanides or sulfides. Acids should never be put into steel containers. Water-reactive, strong acids such as organic acid halides, organic acid anhydrides, inorganic acid anhydrides, and strong acidic salts must be kept apart from both alkalis and water. Oxidizing agents must be kept apart from reducing agents and organic compounds. Water-reactive agents must be stored apart from water, aqueous solutions, and acids. Air-reactive materials must be packed in containers that are sealed off from the atmosphere. Explosive and shock-sensitive materials present risks that require special handling.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Retrieval of Transuranic Drums with Deflagration to Detonation Potential from a Vault at the Oak Ridge National Laboratory - 20207

During the final phase of a long-term project, a total of 80 TRU waste drums at ORNL have been retrieved from a below-grade storage vault in which they have remained untouched for about 25 years. The drums, which house material generated from past ORNL operations, are now stored and ready for eventual transfer to the Transuranic Waste Processing Center, located on Highway 95 in Oak Ridge. The Transuranic Waste Processing Center will process and ship the drums to WIPP in Carlsbad, New Mexico, for disposal. Workers lifted the drums one-by-one from the underground storage vault with a 110-ton crane over a one-week period in October 2017. Although the physical work was performed in one week, years of detailed planning preceded the fieldwork. A key obstacle needed to be overcome before the project could safely proceed. This obstacle involved the need to ensure controls were developed for the Deflagration to Detonation Transition (DDT) potential. DDT refers to a phenomenon in ignitable mixtures of a flammable gas and air (or oxygen) when a sudden transition takes place from a deflagration type of combustion to a detonation type of explosion. Basically, these drums, under certain circumstances, have a potential to detonate and overpressurize. Extensive analysis of the drums verified that detonation, while unlikely, was possible. Of the total 103 drums that were retrieved or moved in preparation for retrieval, 18 had DDT potential. Extensive security requirements were also necessary, due to the nature of the material, and this added another layer of complexity to the fieldwork. Prior to retrieving the drums from the underground storage vault, space needed to be made for them within the various above-grade facilities. Inventory and security requirements dictated what could be stored where, and the first step of the field work involved a number of drum movements in preparation for storage of the newly retrieved drums. This effort involved moving 21 drums already in the storage area to facilitate the addition of the newly retrieved ones. The retrieval project required a total of 107 critical lifts using the 110-ton crane, under 3 critical lift plans. The TRU waste storage facility is a Category 2 Nuclear Facility and is managed under an extensive Documented Safety Analysis. The TRU retrieval project was conducted in accordance with DOE Order 425.1D and included three Implementation Verification Reviews, two Management Assessments, one Readiness Assessment, and seven mock-ups in preparation for the field work. Despite being stored for 25 years, the drums were in good shape due to the integrity of the facilities where the drums were stored. Of the 80 drums that were retrieved, 12 were placed in special containers called 'overpacks' based on their content. The drums were removed safely and without incident, thanks to a lot of preparation and the efforts of a highly skilled team. Lessons learned through the DDT control development and the drum retrieval and overpacking process could be beneficial to others having to manage similar drums of TRU waste. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Using Robotic Manipulators for Radioactive Waste Inspection - 20090

The global nuclear industry has a growing volume of nuclear waste which needs to be scanned, sorted according to its activity and material type, then processed into the correct waste packages for long term storage and disposal. It is vital that there is a detailed understanding of the waste inventory stored in long term waste containers, as knowledge of their contents could predict or prevent any adverse effects in storage. The numerous 'scan and sort' tables which are currently used at many different facilities around the world to sort waste into their correct containers are human operated and require very slow gamma scanning procedures combined with educated guesswork to manually sort the waste. This often leads to excessive conservatisms, with placement of lower activity wastes in higher activity containers, which in turn costs significantly more to store. In the United Kingdom it costs UK Pounds 46 k per cubic meter to store intermediate level waste compared to just UK Pound 2.9 k per cubic meter to store low level waste according to a 2008 Department of Energy and Climate Change report in the UK. A proposed solution to this problem, is the use of a robotic manipulator to automatically inspect the 'scan and sort table' in order to produce an accurate 3D model of the table's waste contents and attach an overlaid radiation map. The radiation map contains spectrometry data and can in consequence be used to distinguish and locate specific radioisotopes. The 3D model should be as accurate as possible in order to allow for a second robot arm with an attached gripper to grasp the objects and place them into their designated long-term storage container. Various scanning procedures are explored in this study including basic raster scanning, adaptive raster scanning and point sampling. The optimal solution will in practice be defined by the required application and activity level of the wastes being inspected. The results presented in this study indicate that it is possible to produce a centimeter accurate 3D model of a mixed assortment of components on a nuclear waste 'scan and sort' table. In addition, it was shown that the waste objects emitting radiation could be accurately identified and located, with an overlaid radiation map. This study is applicable across the nuclear waste management sector. Many of the ideas and concepts developed in this study are applicable in other decommissioning settings for example, dismantling of legacy gloveboxes or routine inspection of nuclear waste packages in storage. (authors)

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

Total Integrated Dose to Spent Fuel Pool Level Instrument Probes and Refueling Floor Personnel Dose Rates Meeting NRC Order EA-12-051 for a Typical Boiling Water Reactor - Paper 22

In reviewing the events after the March 11, 2011 Fukushima accident, the Nuclear Regulatory Commission (NRC) issued Order EA-12-051, Reference 1, requiring new reliable spent fuel pool (SFP) instrumentation for pool level determination under beyond design basis event conditions. In 2013, URS prepared a calculation for a typical Boiling Water Reactor to determine the Total Integrated Dose (TID) potentially to be imparted to radiation sensitive underwater probes (one primary and one backup) planned for meeting this Order. Additionally, URS calculated the radiological dose rates to an individual standing on the refueling floor for various SFP fill levels. NEI 12-02, Reference 2, defines the levels at which SFP condition identification is necessary as the following: Level 1 - level that is adequate to support operation of the normal fuel pool cooling system. Level 1 represents the higher of the following two points: - The level at which reliable suction loss occurs due to uncovering of the coolant inlet pipe, weir or vacuum breaker (depending on the design), or - The level at which the water height, assuming saturated conditions, above the centerline of the cooling pump suction provides the required net positive suction head specified by the pump manufacturer or engineering analysis. Level 2 - level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck. Level 2 is based on either of the following: - 10 feet (+/- 1 foot) above the highest point of any fuel rack seated in the spent fuel pools, or - A designated level that provides adequate radiation shielding to maintain personnel radiological dose levels within acceptable limits while performing local operations in the vicinity of the pool. This level shall be based on either plant-specific or appropriate generic shielding calculations, considering the emergency conditions that may apply at the time and the scope of necessary local operations, including installation of portable SFP instrument channel components. Level 3 - level where fuel remains covered and actions to implement make-up water addition should no longer be deferred. Level 3 corresponds nominally (i.e. +/- 1 foot) to the highest point of any fuel rack seated in the spent fuel pool. The URS calculation did not define these levels or determine instrumentation setpoints. However, it determined refueling floor dose rates that could be used as a tool to determine the appropriate water depths for these levels. In determining the TID for the SFP probes and the radiological dose rates to the individual, the following radioactive sources were considered: 1. Spent Nuclear Fuel (SNF) stored in the spent fuel storage racks, 2. Local Power Range Monitors (LPRMs) stored above the top of the racks on hangers on the SFP walls 3. Radioactive material laden filters stored on platforms on top of the racks, and 4. Control Rod Blades (CRBs) stored above the top of the racks on hangers on the SFP walls. The TID for the SFP probes includes both the dose from normal operation (pool full of water) for the remaining plant life (an assumed 40 years, conservatively including an allowance for lifetime extension SFP loading) and, consistent with the NEI 12-02 guidelines, Reference 2, the assumed drop in elevation of the pool water level to the top of the spent fuel racks for a 6 hour period. The acceptable TID for the radiation-sensitive probe as defined by the manufacturer was 2x10{sup 9} rad. The individual on the pool deck was assumed to be standing on the edge of the pool with an unobstructed view of its contents, as a worst case scenario. Dose rates to this individual determined from direct shine from the same 4 sources were determined for SFP water levels varying from 1 ft above the SNF racks to 21 ft above the racks in one foot increments. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

The experiment filtering tool in the MACSENS workbench - 14341

The MACSENS (outils Moret d'Aide a la Caracterisation et aux etudes de SENSibilite) workbench is dedicated to the advanced validation work for criticality studies. The aim of this kind of tool is to determine the experiments from a validation database being the 'closest' to an application case and, from there, to evaluate a calculation bias with a related uncertainty. To achieve this goal, the MACSENS workbench relies on different kind of information stored in various databases depending on the origin of the data. Two main kinds of data can be identified. The first one concerns the data used to describe the experiments of the validation database (e.g. the inputs needed to perform a calculation). These values characterize the chemical and geometrical models. They can be considered as constant, as they do not vary once evaluated, and some of them can be from a non-numerical nature (reflector material for example...). The second kind of data gathers the results obtained with the use of a simulation code, such as reaction rates, fluxes and leakages, sensitivities, k{sub eff}, etc. These values depend on the version of the code and are purely numerical. In the early 2000's, regarding the increasing number of experiments evaluated in the ICSBEP handbook, the OECD/NEA (Organization for Economic Co-operation and Development/Nuclear Energy Agency), under the ICSBEP (International Criticality Safety Benchmark Evaluation Project) working group direction, decided to store such information in a database and to provide a software, called DICE (Database for the International Criticality safety benchmark Evaluation project), allowing users to search easily criticality experiments of interest. Up to now, more than 4400 critical or subcritical configurations have been compiled by ICSBEP, described and stored in the DICE database. Over time, a validation work has been done on the data, with the participation of IRSN, and a quality assurance policy has been defined. For these reasons, the DICE database is highly reliable. These features were fundamental for the MACSENS developers, who were looking for a database containing descriptive data of the experiments included in the MORET code validation database. The paper describes the solution chosen by IRSN to take advantage of the DICE database content and to overcome the DICE limits. It also explains the quality assurance policy applied to guaranty the reliability of the stored data. It will also discuss the use of the experiment-filtering module of the MACSENS workbench to select a set of experiments and to examine numerical parameter trends for an application case. (authors)

42 ENGINEERING↗
DOE OSTI2024

Systems and methods for geothermal energy storage

A method and system may be used for storing energy in a geothermal system and recovering both the stored energy as well as thermal energy on demand. The geothermal system may include injection and production wells that are hydraulically coupled in a geothermal energy reservoir that behaves as a confined reservoir system with thermal energy transferring to fluid injected into the injection well and removed via the production well. Injection flow rate, injection pressure, production flow rate, production backpressure, or fluid residence time may be managed to control energy consumption or energy generation profiles of the geothermal system. During an energy storage mode, the injection flow rate exceeds the production flow rate thereby storing energy in the geothermal reservoir. During an energy recovery mode, production backpressure is reduced thereby releasing the stored energy and electricity is generated by removing the thermal energy from the fluid in a heat engine.

Norbeck, Jack↗
DOE OSTI2025

Synergistic heat pumped thermal storage and flexible carbon capture system

A power plant system is disclosed. The power plant system includes a combustor configured, a turbine configured to generate electricity, a heat exchanger and a steam turbine, a carbon capture system configured to remove at least a portion of carbon-based gasses from the flue gas downstream from the heat recovery steam generator, and a thermal storage system including a hot storage unit configured to store thermal energy at a hot temperature, the hot temperature greater than ambient temperature. The power plant is configured to operate in at least a first mode for storing thermal energy in the thermal storage system and a second mode for releasing the stored thermal energy from the thermal storage system and during the second mode, heat stored in the hot storage unit is transferred to the carbon capture system.

Bandhauer, Todd M.↗
DOE OSTI2019

Robust control of the current profile and plasma energy in EAST

Integrated control of the toroidal current density profile, or alternatively the q-profile, and plasma stored energy is essential to achieve advanced plasma scenarios characterized by high plasma confinement, magnetohydrodynamics stability, and noninductively driven plasma current. The q-profile evolution is closely related to the evolution of the poloidal magnetic flux profile, whose dynamics is modeled by a nonlinear partial differential equation (PDE) referred to as the magnetic-flux diffusion equation (MDE). The MDE prediction depends heavily on the chosen models for the electron temperature, plasma resistivity, and non-inductive current drives. To aid control synthesis, control-oriented models for these plasma quantities are necessary to make the problem tractable. However, a relatively large deviation between the predictions by these control-oriented models and experimental data is not uncommon. For this reason, the electron temperature, plasma resistivity, and non-inductive current drives are modeled for control synthesis in this work as the product of an “uncertain” reference profile and a nonlinear function of the different auxiliary heating and current-drive (H&CD) source powers and the total plasma current. The uncertainties are quantified in such a way that the family of models arising from the modeling process is able to capture the q-profile and plasma stored energy dynamics from a typical EAST shot. A control-oriented nonlinear PDE model is developed by combining the MDE with the “uncertain” models for the electron temperature, plasma resistivity, and non-inductive current drives. This model is then rewritten into a control framework to design a controller that is robust against the modeled uncertainties. The resulting controller utilizes EAST's H&CD powers and total plasma current to regulate the q profile and plasma stored energy even when mismatches between modeled and actual dynamics are present. The effectiveness of the controller is demonstrated through nonlinear simulations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Growth kinetics of Al–Fe intermetallic compounds during annealing treatment of friction stir lap welds

In this study, we explored the growth kinetics of the Al–Fe intermetallic (IM) layer at the joint interface of the St-12/Al-5083 friction stir lap welds during post-weld annealing treatment at 350, 400 and 450 °C for 30 to 180 min. Optical microscope (OM), field emission gun scanning electron microscope (FEG-SEM) and transmission electron microscope (TEM) were employed to investigate the structure of the weld zone. The thickness and composition of the IM layers were evaluated using image analysis system and electron back-scatter diffraction (EBSD), respectively. Moreover, kernel average misorientation (KAM) analysis was performed to evaluate the level of stored energy in the as-welded state. The results showed that the growth kinetics of the IM layer was not governed by a parabolic diffusion law. Presence of the IM compounds as well as high stored energy near the joint interface of the as-welded sample was recognized to be the origin of the observed deviation from the parabolic diffusion law. - Highlights: • This work provided a new insight into growth kinetics of Al–Fe IM thickness. • The growth kinetics of IM layer was not governed by a parabolic diffusion law. • IM near the joint interface was the origin of deviation from the parabolic law. • High stored energy at joint interface was origin of deviation from parabolic law.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Microstructure characterization based on the type of deformed grains in cold-rolled, Cu-added, bake-hardenable steel

The electron backscatter diffraction technique has been used to characterize the microstructure of deformed grains in cold-rolled, Cu-added, bake-hardenable steel. A new scheme based on the kind and number of average orientations, as determined from a unique grain map of the deformed grains, was developed in order to classify deformed grains by type. The α-fiber components, γ-fiber components and random orientations, those which could not be assigned to either γ-fiber or α-fiber components, were used to define the average orientation of unique grains within individual deformed grains. The microstructures of deformed grains in as-rolled specimens were analyzed based on the Taylor factor, stored energy, and misorientation. The relative levels and distributions of the Taylor factor, the stored energy and the misorientation were examined in terms of the types of deformed grains. - Highlights: • We characterized the microstructure of Cu-added BH steel using EBSD. • A new scheme was developed in order to classify deformed grains by type. • Stored energy and misorientation are strongly dependent on the type of deformed grains. • Microstructure was examined in terms of the types of deformed grains.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Quantifying the inter- and intra-granular stresses and their effects on deformation behaviors for the polycrystalline hexagonal beryllium by neutron diffraction

Highlights: • The inter- and intra-granular stresses are quantified for the polycrystalline hexagonal metal beryllium. • Composite behavior of the internal stresses is revealed in the polycrystalline beryllium. • Those internal stresses show the obvious grain-orientation dependence. • Those stresses can be significantly tuned by the temperature and strain rate effects. • The dislocation slip and twinning are quantified together with the inter- and intra-granular stored energies. - Abstract: The internal stresses developing at different length scales play the important role in materials performance. This importance has been well recognized for the cubic metals, however, not for the hexagonal close-packed (hcp) ones. The present work investigates the intergranular and intragranular stresses in the polycrystalline hexagonal beryllium by neutron diffraction. The correlation between those internal stresses and the deformation mechanism is revealed, and the strain rate and temperature effects on those stresses are presented. A composite behavior of intergranular and intragranular stresses is observed, which can be significantly tuned by the temperature and strain rate effects. The intergranular stress and stored energy are large for the soft-oriented grains (SOGs), while the intragranular stress and stored energy are large for the hard-oriented ones (HOGs). Both of those stresses are significantly high for the dynamic strain rate while they are relatively low for the elevated temperature. The large internal stresses in the dynamically compressed specimen should be attributed to the large number of twins and high density of dislocations. Different HOGs-SOGs configurations are proposed on the light of those observations, which might facilitate the properties controlling and extend the application scope for the hcp metal beryllium.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Integration of a Wigner effect-based energy storage system with an advanced nuclear reactor

In this work, an innovative energy storage concept based on the purposeful creation of defects in crystalline material by neutron irradiation is presented. Lattice defects are generated when heavy particles collide with the atoms in a crystal structure, i.e., if the incoming particles have enough energy, recoil atoms are displaced from their initial lattice sites. Most of the displaced atoms will eventually combine with nearby vacancies, but some of them will come to rest in non-ideal locations. The energy held by displaced atoms is called Wigner energy. Lattice defects can migrate and form clusters, and the Wigner energy can be released from these groupings if sufficient activation energy is provided. In the nuclear industry, this effect is well-known since it represented an issue for graphite-moderated reactors. This work presents the conceptual design of an engineering system that exploits this physical process to store the energy of neutrons in advanced reactor concepts. In the first part of the paper, the theoretical performance of an energy storage system based on the Wigner effect is described. Given the lattice properties and the compatibility with the harsh reactor environment, graphite was selected as the candidate material for the irradiation targets. Both experimental data and molecular dynamics simulations confirmed that this system can achieve performance comparable with state-of-the-art batteries in terms of stored energy density. In the second part of the paper, the engineering challenges of this innovative technology and the proposed solutions are described. After defining the optimal irradiation conditions, the different steps of the operation of the proposed energy system (from energy storing to energy harvesting) were defined. Finally, the integration of this concept with advanced reactor designs, i.e., a Sodium-cooled Fast Reactor and a Molten Salt-cooled Reactor, was investigated and the corresponding performance was evaluated.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Improvement in the statistical operation of a Blumlein pulse forming line in bipolar pulse mode

The paper presents the results of studies on shot-to-shot performance of a water Blumlein pulse forming line of 1–1.2 kJ of stored energy. The experiments were carried using the TEMP-4M pulsed ion beam accelerator during its operation in both unipolar pulse mode (150 ns, 250–300 kV) and bipolar-pulse mode with the first negative (300–600 ns, 100–150 kV) followed by a second positive (120 ns, 250–300 kV) pulse. The analysis was carried out for two cases when the Blumlein was terminated with a resistive load and with a self-magnetically insulated ion diode. It was found that in bipolar pulse mode the shot-to-shot variation in breakdown voltage of a preliminary spark gap is small, the standard deviation (1σ) does not exceed 2%. At the same time, the shot-to-shot variation in the breakdown voltage of the main spark gap in both bipolar-pulse and unipolar pulse mode is 3–4 times higher than that for the preliminary spark gap. To improve the statistical performance of the main spark gap we changed the regime of its operation from a self-triggered mode to an externally triggered mode. In the new arrangement the first voltage pulse at the output of Blumlein was used to trigger the main spark gap. The new trigatron-type regime of the main spark gap operation showed a good stability of breakdown voltage and thus allowed to stabilize the duration of the first pulse. The standard deviation of the breakdown voltage and duration of the first pulse did not exceed 2% for a set of 50 pulses. The externally triggered mode of the main gap operation also allowed for a decrease in the charging voltage of the Blumlein to a 0.9–0.95 of self-breakdown voltage of the main spark gap while the energy stored in Marx generator was decreased from 4 kJ to 2.5 kJ. At the same time the energy stored in Blumlein remained the same.

43 PARTICLE ACCELERATORS↗
DOE OSTI2016

Space-time evolution of ejected plasma for the triggering of gas switch

Ejected plasma has been widely applied to the discharge process of gas spark switches as a trigger technology, and the development process of ejected plasma has a direct and important effect on the discharge characteristics of gas switches. In this paper, both the injection characteristics and space-time evolution of ejected plasma for the triggering of gas spark switch with different stored energies, pulse polarities, and pressures are studied. The discharge characteristics and breakdown process of a gas switch ignited by ejected plasma under different working coefficients are also discussed briefly. The results show that stored energy has significant influence on the characteristics of ejected plasma. With the increase of stored energy, the propulsion mode of ejected plasma in the axial direction transforms from “plasmoid” to “plasma flow,” and the distribution of the ejected plasma goes through “cloud,” “core-cloud,” and “branch” in sequence. The velocity of ejected plasma under negative pulse polarity is obviously higher than that under positive pulse polarity, especially at the very beginning time. The radial dimensions of ejected plasma under two kinds of pulse polarities follow the similar varying pattern over time, which increase first and then decrease, assuming an inverted “U”-shaped curve. With the increase of pressure, the velocity of ejected plasma significantly decreases and the “branch” channels droop earlier. Applying the ejected plasma to the triggering of a gas switch, the switch can be triggered reliably in a much wide working coefficient range of 10%–90%. With the increase of working coefficient, the breakdown process of the switch translates from slow working mode to fast working mode, and the delay time reduces from tens of μs to hundreds of ns.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗