SEARCH · Engineering Papers
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.
Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.
According to one embodiment, a composite product includes hexagonal boron nitride (hBN), and a plurality of cubic boron nitride (cBN) particles, wherein the plurality of cBN particles are dispersed in a matrix of the hBN. According to another embodiment, a composite product includes a plurality of cBN particles, and one or more borate-containing binders.
A rechargable magnesium battery having an non-aqueous electrolyte is provided. The properties of the electrolyte include high conductivity, high Coulombic efficiency, and an electrochemical window that can exceed 3.5 V vs. Mg/Mg.sup.+2. The use of the electrolyte promotes the electrochemical deposition and dissolution of Mg without the use of any Grignard reagents, other organometallic materials, tetraphenyl borate, or tetrachloroaluminate derived anions. Other Mg-containing electrolyte systems that are expected to be suitable for use in secondary batteries are also described.
According to one embodiment, a composite product includes: a matrix material including hexagonal boron nitride and one or more borate binders; and a plurality of cubic boron nitride particles dispersed in the matrix material. According to another embodiment, a composite product includes: a matrix material including hexagonal boron nitride and amorphous boron nitride; and a plurality of cubic boron nitride particles dispersed in the matrix material.
Methods of converting lignocellulosic derivatives form 5-hydroxymethyl furfural (HMF) in ionic liquids are provided herein. Glucose is enzymatically converted to fructose by isomerization with glucose isomerase and a borate salt. Conversion of fructose to HMF includes dehydration of fructose using ionic liquids or ionic liquids with an acid catalyst to achieve high yield of HMF.
To advance the current understanding of heat transfer between nuclear fuel rod cladding and the surrounding coolant under transient scenarios involving critical heat flux (CHF), the CHF-Static Environment Rodlet Transient Test Apparatus (CHFSERTTA), an in-pile experiment, was developed for and irradiated in the Transient Reactor Test (TREAT) facility at Idaho National Laboratory. The CHF-SERTTA experiment consists of a hollow, borated stainless steel rod submerged in a static water pool. The rod, heated via the (n, a) reaction in boron-10, was designed for investigating in-pile transient heat transfer between the rod surface and the coolant, avoiding the complexity of including the heat transfer between the fuel, gas gap, and cladding in typical light-water reactor (LWR) fuel rods. The instrumentation package for CHF-SERTTA focused on capturing those data that best describe the heat transfer behaviour between the rod and coolant. This paper outlines the CHFSERTTA experiment design and presents the initial results of the experiment campaign.
This White Paper brings together our previous studies on a Muon Collider (MC) and presents a design concept of the 6 TeV MC optics, the superconducting (SC) magnets, and a preliminary analysis of the protection system to reduce magnet radiation loads as well as particle backgrounds in the detector. The SC magnets and detector protection considerations impose strict limitations on the lattice choice, hence the design of the collider optics, magnets and Machine Detector Interface (MDI) are closely intertwined. As a first approximation we use the Interaction Region (IR) design with beta-star=3 mm, whereas for the arcs we re-scale the arc cell design of the 3 TeV MC. Traditional cos-theta coil geometry and Nb3Sn superconductor were used to provide field maps for the analysis and optimization of the arc lattice and IR design, as well as for studies of beam dynamics and magnet protection against radiation. The stress management in the coil will be needed to avoid large degradation or even damage of the brittle SC coils. In the assumed IR designs, the dipoles close to the Interaction Point (IP) and tungsten masks in each IR (to protect magnets) help reducing background particle fluxes in the detector by a substantial factor. The tungsten nozzles in the 6 to 600 cm region from the IP, assisted by the detector solenoid field, trap most of the decay electrons created close to the IP as well as most of the incoherent electron-positron pairs generated in the IP. With sophisticated tungsten, iron, concrete and borated polyethylene shielding in the MDI region, the total reduction of background loads by more than three orders of magnitude can be achieved.
A method for obtaining solid-state metal borohydrides without toxic precursors and expensive solvents includes dry mixing of metal hydrides and metal polyhydro-closo-borate starting materials. High pressure and heating is also used in the method. These materials can be used for hydrogen storage, general reducing agents, organic synthesis, wastewater treatment, and paper pulp bleaching.
Electrolytes for lithium ion batteries with carbon-based, silicon-based, or carbon- and silicon-based anodes include a lithium salt; a nonaqueous solvent comprising at least one of the following components: (i) an ester, (ii) a sulfur-containing solvent, (iii) a phosphorus-containing solvent, (iv) an ether, (v) a nitrile, or any combination thereof, wherein the lithium salt is soluble in the solvent; a diluent comprising a fluoroalkyl ether, a fluorinated orthoformate, a fluorinated carbonate, a fluorinated borate, or a combination thereof, wherein the lithium salt has a solubility in the diluent at least 10 times less than a solubility of the lithium salt in the solvent; and an additive having a different composition than the lithium salt, a different composition than the solvent, and a different composition than the diluent.
The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor. This paper is part of a series of four describing the design and safety features of the Westinghouse SMR. This paper focuses in particular upon the passive safety features and the safety system response of the Westinghouse SMR. The Westinghouse SMR design incorporates many features to minimize the effects of, and in some cases eliminates the possibility of postulated accidents. The small size of the reactor and the low power density limits the potential consequences of an accident relative to a large plant. The integral design eliminates large loop piping, which significantly reduces the flow area of postulated loss of coolant accidents (LOCAs). The Westinghouse SMR containment is a high-pressure, compact design that normally operates at a partial vacuum. This facilitates heat removal from the containment during LOCA events. The containment is submerged in water which also aides the heat removal and provides an additional radionuclide filter. The Westinghouse SMR safety system design is passive, is based largely on the passive safety systems used in the AP1000{sup R} reactor, and provides mitigation of all design basis accidents without the need for AC electrical power for a period of seven days. Frequent faults, such as reactivity insertion events and loss of power events, are protected by first shutting down the nuclear reaction by inserting control rods, then providing cold, borated water through a passive, buoyancy-driven flow. Decay heat removal is provided using a layered approach that includes the passive removal of heat by the steam drum and independent passive heat removal system that transfers heat from the primary system to the environment. Less frequent faults such as loss of coolant accidents are mitigated by passive injection of a large quantity of water that is readily available inside containment. An automatic depressurization system is used to reduce the reactor pressure in a controlled manner to facilitate the passive injection. Long-term decay heat removal is accomplished using the passive heat removal systems augmented by heat transfer through the containment vessel to the environment. The passive injection systems are designed so that the fuel remains covered and effectively cooled throughout the event. Like during the frequent faults, the passive systems provide effective cooling without the need for ac power for seven days following the accident. Connections are available to add additional water to indefinitely cool the plant. The response of the safety systems of the Westinghouse SMR to various initiating faults has been examined. Among them, two accidents; an extended station blackout event, and a LOCA event have been evaluated to demonstrate how the plant will remain safe in the unlikely event that either should occur. (authors)
In 2001 GEH applied best estimate methods combined with a statistical methodology to determine upper bound limits for key licensing parameters for anticipated operation occurrence (AOO) transient and anticipated transients without scram (ATWS) overpressure analyses for operating Boiling Water Reactors (BWRs). The methodology was subsequently extended for ESBWR AOO, ATWS, loss of coolant, and stability analyses. GEH is extending the methodology to long-term ATWS analyses for the operating BWRs. A long-term ATWS scenario uses injection of borated water to achieve reactor shutdown. Predicting the mixing and transport of boron is important for calculating the impact on the key licensing parameters. For the many operating BWRs where the denser boron solution is injected into the lower plenum, stratification may occur, delaying boron transport to the core region. CFD modeling can be used to model the stratification and mixing of the boron solution, but such calculations are extremely computer intensive and not cost effective; therefore, a more-empirical approach supported by a theoretical scaling of the dominant phenomena and backed by test data and benchmark calculations is used. The paper presents the TRACG lower plenum boron transport model qualification effort. The scaling basis used to implement the TRACG boron transport model for BWR applications is discussed. (authors)
Electrolytes for lithium ion batteries with carbon-based, silicon-based, or carbon- and silicon-based anodes include a lithium salt; a nonaqueous solvent comprising at least one of the following components: (i) an ester, (ii) a sulfur-containing solvent, (iii) a phosphorus-containing solvent, (iv) an ether, (v) a nitrile, or any combination thereof, wherein the lithium salt is soluble in the solvent; a diluent comprising a fluoroalkyl ether, a fluorinated orthoformate, a fluorinated carbonate, a fluorinated borate, a fluorinated phosphate, a fluorinated phosphite, or any combination thereof, wherein the lithium salt has a solubility in the diluent at least 10 times less than a solubility of the lithium salt in the solvent; and an additive having a different composition than the lithium salt, a different composition than the solvent, and a different composition than the diluent. In some electrolytes, the nonaqueous solvent comprises an ester.
Sample glasses have been made using SB6 high level waste (HLW) simulant (high in both Al and Fe) with 12 different frit compositions at a constant waste loading of 36 wt.%. As follows from X-ray diffraction (XRD) and optical and scanning electron microscopy (SEM) data, all the samples are composed of primarily glass and minor concentration of spinel phases which form both isometric grains and fine cubic (∼1 μm) crystals. Infrared (IR) spectra of all the glasses within the range of 400-1600 cm-1 consist of the bands due to stretching and bending modes in silicon-oxygen, boron-oxygen, aluminum-oxygen and iron-oxygen structural groups. Raman spectra showed that for the spectra of all the glasses within the range of 850-1200 cm{sup -1} the best fit is achieved by suggestion of overlapping of three major components with maxima at 911-936 cm{sup -1}, 988-996 cm{sup -1} and 1020-1045 cm{sup -1}. The structural network is primarily composed of metasilicate chains and possibly rings with embedded AlO{sub 4} and FeO{sub 4} tetrahedra. Major BO{sub 4} tetrahedra and BO{sub 3} triangles form complex borate units and are present as separate constituents. (authors)
SIAL{sup R} is a proven, safe, cost-effective and licensed waste solidification technology that has been used successfully for 20 years in Slovakia and the Czech Republic for processing of various problematic radioactive waste streams. This abstract presents and summarises the most interesting performance records of the SIAL{sup R} geopolymer solidification technology, the applied mobile devices as well as the characteristics of final products. Radioactive waste solidification into matrix SIAL{sup R} is based on polycondensation of inorganic aluminosilicate components with chemical composition similar to average cement. The main components of matrix SIAL{sup R} are natural inorganic raw materials containing SiO{sub 2}, Al{sub 2}O{sub 3}, CaO, MgO, Na{sub 2}O, K{sub 2}O, Fe{sub 2}O{sub 3}, and TiO{sub 2}. The treatment/solidification process is modified for individual radioactive streams with different compositions. To determine the most appropriate composition, the SIAL{sup R} application technology is laboratory tested on real samples of each individual waste stream identified for solidification. Then an appropriate mixture of aluminosilicate composition and other inorganic compounds is defined for a given solidification application. When determining the most appropriate mixture, other requirements are taken into consideration: maximum dose rate on the drum with solidified product, mass, volume, alpha radionuclide content in the final product, need for a waste's dewatering before solidification, dry matter content in the solidification product as well as other specific waste acceptance criteria (i.e., compressive strength, leachability, diffusion coefficients, radiation stability, biodegradability of matrix, distribution and partition coefficients, influence of frost and water presence to the product stability including long term physical stability in water, dustiness, flammability, time dependence of compression strength and water reduction, product homogeneity and also radiolytic G-value for hydrogen release, GH{sub 2}). The Amec Foster Wheeler baseline approach to implementing the SIAL{sup R} waste treatment process is structured on three steps - waste characterization, waste retrieval and pre-treatment and at last the solidification of pre-treated waste itself. The equipment used for the solidification process is relatively small, lightweight and designed for the specific application on site, directly 'in situ' or local to the tanks storing the sludge's/resins/other waste streams. This mitigates the problems normally experienced by transferring sticky, tight waste through long pipes. The equipment is normally 'tailor-made' to take into account the real space conditions and other existing facility requirements. At Bohunice and Dukovany sites, the facilities for encapsulation were designed to occupy existing areas of the plants in close proximity to the tanks where the waste was being stored. This allowed for the removal of the waste using Remote Operated Vehicles and for the waste to be immediately treated on site without the need for transportation thus saving both, time and resource. To date, SIAL{sup R} has been used successfully to immobilize approximately 2 x 10{sup 6} kg of waste which includes sludge, resin, crystalline borates and contaminated organic waste from Bohunice and Mochovce, Slovakia and approximately 300 m{sup 3} of spent ion exchange resins and sludge from tanks on site at the Dukovany NPP (Nuclear Power Plant) in the Czech Republic. The technology now has a track record of over 20 years which includes on-going research and development. The International missions WANO (the World Association of Nuclear Operators) and OSART (the Operational Safety Review Team) evaluated the SIAL{sup R} matrix technology at Dukovany NPP as an example of good practice. (authors)
On March 11, 2011 a 9.0 magnitude earthquake occurred off the coast of Japan creating a tsunami, which together severely damaged northeastern Japan including the Fukushima Dai-ichi nuclear power plant site. In response to this event, the United States (U.S.) Nuclear Regulatory Commission (NRC) issued Order EA-12-049, 'Issuance of Order to Modify Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events'. In order to meet the guidance of the NRC Order EA-12- 049, the a central U.S. nuclear generating station (Plant) was required to demonstrate that the currently installed Borated Water Storage Tank (BWST), Refueling Water Tank (RWT), and Qualified Condensate Storage Tank (QCST) could provide the necessary capacity following a Beyond Design Basis External Event (BDBEE). These tanks were not originally designed to meet current site design basis requirements for tornado wind and missile impacts. To save the Plant from having to build costly new storage tanks off site, the existing tanks were credited as robust for tornado missile impact through advanced finite element impact analyses using the ANSYS Autodyn solver and fluid-structure interaction (FSI). This procedure is the first known NRC acceptance of advanced analysis and FSI for to satisfy the FLEX tornado missile impact criteria. Extensive verification and validation (V and V) was performed to verify perforation of steel plates which included FSI. These V and V analyses were confirmed against large databases of experimental and theoretical results to ensure accurate modeling and results. The use of this advanced analytical method benefitted the Plant in several ways. A new storage tank was not required to be built, saving not only initial construction costs but also ongoing maintenance costs and monitoring activities. By determining specific vulnerable components of each tank a set of modifications and protection schemes were developed. These modifications were much quicker to implement saving the Plant on downtime costs. (authors)
The Experimental Breeder Reactor II (EBR-II) was selected as a candidate for a reactor physics benchmark evaluation due to its unique features and capabilities. In connection with the benchmark evaluation, the goal of this work was to perform a sensitivity analysis. The EBR-II was a sodium cooled, pool type, metal fueled, fast spectrum reactor. EBR-II operated from 1964 through 1994. It had a maximum heat output of 62.5 MW which could produce 20 MW of electricity. The design of EBR-II consisted of 637 hexagonal shaped, removable assemblies. The assemblies were divided into three regions: the core, an inner blanket, and an outer blanket. The core region contained assemblies with 91 fuel rods in each. The fuel was enriched uranium metal and was clad within stainless steel. The core region also included two safety assemblies and eight control assemblies with the difference being that these assemblies only contained 61 fuel rods. Additional assemblies in the core region included stainless steel dummies, half worth drivers, and experimental/instrumentation assemblies. The inner blanket region initially consisted of depleted uranium for fuel breeding, but was replaced, after proving the concept of breeding, with stainless steel reflectors for the purpose of irradiation. The outer blanket region consisted of depleted uranium. Figure 1 shows a cross section view of the EBR-II core layout. Each assembly was approximately 234 cm long with the uranium fuel section being 33 cm long. Above and below the fuel were stainless steel neutron reflectors. Additionally, orifices at the bottom of the assembly allowed for sodium coolant to flow upward, with larger holes in the subassemblies at the center where the greatest heat was produced. The core of EBR-II was enclosed by a stainless steel reactor vessel which was comprised of a radial shell, a lower grid plenum structure, and an upper reactor vessel cover that served as a neutron shield. Surrounding the reactor vessel radially were layers of graphite and borated graphite shielding. The entire assembly is submerged 3 meters in a pool of sodium. The pool contained up to 337,000 liters of sodium which was forced through the core and blanket regions using two centrifugal pumps that were also submerged within the sodium pool. The design of EBR-II included many safety features. The compatibility between the sodium and stainless steel prevented corrosion and therefore the risk of a radioactive release due to corrosion. The sodium coolant was kept at near atmospheric pressure which made any leak more controllable. The metal fuel, cladding, and coolant allowed for heat to be removed easily. The large volume of sodium allowed for heat to be absorbed without problems occurring. Lastly, the reactor would expand as temperature increased which would result in greater neutron leakage. This extra leakage caused a large negative temperature reactivity feedback which automatically shut the reactor down in case of an accident. EBR-II had unique capabilities which were proven during the Integral Fast Reactor initiative. This program lasted from 1984 to 1994. The goal of this program was to solve many of the perceived problems with nuclear energy through scientific means. The safety features of EBR-II were put to the test in April 1986. The tests simulated loss of reactor coolant flow and loss of a heat sink. For both of these tests, the reactor was at 100% power and was not scrammed. During both tests, the reactor shut itself down and achieved a safe steady state temperature all due to the design of the reactor. It was because of these tests that EBR-II was chosen for a reactor physics benchmark evaluation. (authors)
Used nuclear fuel contains fissionable materials ({sup 235}U, {sup 239}Pu, {sup 241}Pu, etc.). To prevent nuclear criticality in used fuel storage, transportation, and disposal, neutron-absorbing materials (or neutron poisons, such as borated stainless steel, Boral{sup TM}, Metamic{sup TM}, Ni-Gd, etc.) would have to be applied. Corrosion-resistant, iron-based structural-amorphous metals (SAMs) have been tested to determine their relative corrosion resistance. Many of these materials can be applied as coatings with advanced High Velocity Oxy-Fuel (HVOF) thermal-spray technology. SAM2X5, SAM1651, and other SAMs are amorphous-metal composite alloys that have been identified as having outstanding corrosion resistance. Because of its high boron content, SAM2X5 can be applied as the neutron-absorbing coatings to the metallic support structure for criticality-safety controls of used fuel in racks in wet storage pool and baskets inside the storage containers, the transportation cask, and eventually the disposal containers in repository disposal. Research and experiment conducted at Lawrence Livermore National Laboratory indicated that the high boron-containing SAM2X5 coating could be an effective criticality control material for used fuel management. The neutron irradiation experiments and the neutron transmission measurements conducted at McClellan Nuclear Radiation Center indicated that extensive fast neutron irradiation did not change the structure of the amorphous SAM2X5 melt-spun ribbons, and SAM2X5 exhibited effective neutron absorbing capability, similar to Boral{sup TM} and Metamic{sup TM}. (authors)
Spent fuel pools (SFPs) were originally designed with low density storage racks that maintained subcriticality based on geometric separation. With the increasing need for storage, due to elimination of reprocessing as a viable alternative, SFPs with high-density racks required neutron absorber materials to control reactivity and maintain criticality safety margins. The Nuclear Energy Institute (NEI) conducted an industry wide survey in 2013 and the survey was updated recently based on additional utility data provided by Electric Power Research Institute (EPRI). The survey results show that about one-quarter of the pools in the United States either do not have neutron absorber material in the SFP or do not credit it even when it is present. Also, about half of the plants in the U.S. use BORAL{sup R} as the neutron absorber material followed by Metamic, Carborundum, Alcan, and borated stainless steel. For neutron absorber monitoring, the majority of the plants rely on a coupon testing program. However, for those plants that never installed coupons or have exhausted their supply of coupons from coupon sampling, in situ measurements at periodic intervals can be performed to determine the status of the neutron absorber. In this paper, an overview of the types of neutron absorber materials used in SFPs, the distribution of absorber material use by material type, types of monitoring programs as part of aging management programs, operational experience to date with different absorber materials based on monitoring programs, and industry initiatives toward addressing long-term performance issues related to neutron absorber material degradation and monitoring are presented. (authors)
The novel structure of lanthanum hydroxyborate La 2 B 2 O 5 (OH) 2 was synthesized by the reaction of partially hydrolyzed lanthanum and boron oxide in a diamond anvil cell under high-pressure/high-temperature (HPHT) conditions of 30 GPa and ~2,400 K. The single-crystal X-ray structure determination of the lanthanum hydroxyborate revealed: P$\overline{3}$c1 , a = 6.555(2) Å, c = 17.485(8) Å, Z = 6, R 1 = 0.056. The three-dimensional structure consists of discrete planar BO 3 groups and three crystallographically different La ions: one is surrounded by 9, one by 10, and one by 12 oxygen anions. The band gap was estimated using ab initio calculations to be 4.64 eV at ambient pressure and 5.26 eV at 30 GPa. The current work describes the novel HPHT lanthanum hydroxyborate with potential application as a deep-ultraviolet birefringent material.