Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “waste package”

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.

At least 793 records · Page 44
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Bare Fiber Bragg Gratings embedded into concrete buffer Supercontainer concept for nuclear waste storage [ANIMMA--2015-IO-337]

Nuclear power plants have been generating electricity for more than 50 years. In Belgium, 55% of the current energy supply comes from nuclear power. Providing for the safe storage of nuclear waste, including spent fuel (SF) and vitrified high level radioactive waste (HLW), remains an important challenge in the life cycle of nuclear fuel. In this context, the Belgian Agency for Radioactive Waste and Enriched Fissile Materials (ONDRAF/NIRAS) is investigating a reference conceptual design called the Supercontainer (SC) for the packaging of SF and HLW. This conceptual design is based on a multiple-barrier system consisting of a hermetically-sealed carbon steel overpack and a surrounding highly-alkaline concrete buffer. The first one is developed to retain the radionuclides. The two main functions of the buffer are (a) to create a high pH environment around the carbon steel overpack in order to passivate the metal surface and so to slow down the corrosion propagation during the thermal phase and (b) to provide a radiological shielding during the construction and the handling of the Supercontainer. A recent test has been performed to investigate the feasibility to construct the SC. This test incorporated several kinds of sensors including Digital Image Correlation (DIC), Acoustic Emission (AE), corrosion sensing techniques and optical fibers with and without fiber Bragg gratings (FBGs). In particular, several single-mode optical fibers with 4 mm long FBGs with different Bragg wavelengths and distributed along the optical fibers were used. For casting and curing condition monitoring, a number of gratings were incorporated inside the concrete buffer during the first stage of construction. Then other sensors were embedded near a heat source installed in the second stage to simulate the effects of heat generated by radioactive waste. The FBGs were designed to measure both temperature and strain effects in the concrete. To discriminate between these effects special packaging was used for some sensors that were installed very close to the unpackaged ones. Sensors placed in plastic tubes have reduced sensitivity to strain, while the ones inserted in metal tubes are only temperature sensitive and their readings can be directly compared with those obtained from thermocouples located nearby. In addition to monitoring temperature and strain behaviour, embedding also had as objective to determine the impact of the high alkaline environment on the silica fibers over a very long time. This article presents the preliminary results obtained with the different FBGs and provides recommendations for future improvement. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

A regulatory perspective on the response function method for determining allowable content of a radioactive materials transportation package or storage cask - 15023

In order to simplify shielding design, there has been an increasingly use of a so-called dose rate response function method for determining the allowable contents in radioactive materials (including spent fuel) transportation package and spent fuel storage cask designs. In essence, this approach uses a pre-calculated dose rate at a given location outside the storage cask or transportation package for a given source inside the cask. It assumes that there is a one-to-one relationship between a source inside the cask and the dose rate outside the cask, i.e., a dose rate response function of a source. Thus, the allowable content is obtained by simply dividing the targeted dose rate by the response function. This method converts the complex particle transport problem into a simple arithmetic calculation. If used correctly and prudently, this approach can greatly simplify the shielding design of casks. However, the use of this method must demonstrate the validity of the assumption, that is, the nuclear physics property of the media through which the particle transverses must be similar to that used in computing the response function. Because of the assumptions and approximations, either implicitly or explicitly, made in the models for dose rate response functions, to assure safety the users should also consider the errors in the models and the bias and uncertainties associated with the code used to calculate the response function. This paper reviewed the advantages and some of the major potential pitfalls of this method from the regulatory perspective. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Apparatus and method for radioactive waste screening

An apparatus and method relating to screening radioactive waste are disclosed for ensuring that at least one calculated parameter for the measurement data of a sample falls within a range between an upper limit and a lower limit prior to the sample being packaged for disposal. The apparatus includes a radiation detector configured for detecting radioactivity and radionuclide content of the of the sample of radioactive waste and generating measurement data in response thereto, and a collimator including at least one aperture to direct a field of view of the radiation detector. The method includes measuring a radioactive content of a sample, and calculating one or more parameters from the radioactive content of the sample.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

The DISTINCTIVE University Consortium: Theme 2: PuO{sub 2} and Fuel Residues - 18132

The safe and secure management of Pu is a matter of serious international concern, with ∼250 t of separated Pu currently stockpiled worldwide. Nearly 50% of this material is in long-term storage in the UK whilst the British government develops its options for final treatment and disposition. Options for this are: (i) reuse as a fuel in modern reactors (e.g. as MOX fuel or within a fast reactor) prior to disposal; (ii) prompt immobilisation for disposal as soon as is practicable; or, (iii) continued long term storage (prior to disposition). The focus of the work presented here is options (ii) and (iii). The R and D needs of both are now pressing: in option (iii) due to it being the current default; in option (ii) because of a comparative lack of R and D on Pu conditioning and packaging due to policy uncertainty as to whether it would be disposed of in a Geologic Disposal Facility (GDF). Addressing these needs is complicated by Pu's high radioactivity, decay heat and radiotoxicity, nuclear safeguard requirements and, for some UK Pu contaminated materials targeted for disposal, poor inventory. Thus, there is also a critical requirement for underpinning research on characterisation methods for Pu bearing materials. In response to these needs, the DISTINCTIVE Consortium (Decommissioning, Immobilisation and Storage Solutions for Nuclear Waste Inventories, a multi-disciplinary collaboration of 11 universities and 3 key industry partners from across the UK's civil nuclear sector) has been seeking to: - Understand how the structure and properties of PuO{sub 2} change with time in the presence of H{sub 2}O. - Understand the roles these processes play in gaseous product evolution from PuO{sub 2} in storage. - Understand radiation-induced amorphization and dissolution kinetics of Pu waste forms. - Develop novel, fast neutron based radiometric methods for the quantification, isotopic composition assessment and remote imaging of Pu-bearing materials. This paper describes recent progress made towards achieving these objectives. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Final Technical Report on Fluoride Salt Study

This report fulfills the Level 3 Milestone No. M3FT-15IN0311022 titled, “Uranium Electrorefining Using Fluoride Salt” within the Department of Energy (DOE) Work Package No. FT-15IN031102 titled, “U/TRU Codeposition and Fluoride Salt Studies – INL”. This work package falls under the Fuel Cycle Research and Development (FCR&D) Materials Recovery and Waste Form Development (MRWFD) Program 1.02.03.11 titled “Domestic Echem”. The electrorefining of uranium in a LiF-NaF-KF, or FLiNaK, salt containing UF 3 has been performed. Several gram-scale electrodeposits of uranium were collected exhibiting a dendritic morphology similar to uranium electrorefining in chloride salts. Additionally, cyclic voltammetry experiments have been performed at 773 K in FLiNaK-UF 4 , FLiNaK-UF 3 , and FLiNaK-UF 3 -LaF 3 salt mixtures to analyze the feasibility of electrorefining uranium in the FLiNaK salt containing lanthanides. Results are promising, showing a 0.48 V potential window between uranium and lanthanum depositions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Features Events and Processes Relevant to DPC Disposal Criticality Analysis

The Department of Energy is evaluating the technical feasibility of disposal of spent nuclear fuel in dual-purpose canisters in various geologies. As part of ongoing research and development, the effect of potential post-closure criticality events on repository performance is being studied. Many different features, events, and processes (FEPs) could affect the potential for criticality or the extent of a criticality event. Additionally, a criticality event could affect other FEPs. This report uses existing lists of FEPs as a starting point to evaluate the FEPs that could affect or be affected by an in- package criticality event. The evaluation indicates that most of the FEPs associated with the waste form, the waste, or the EBS have some effect on post-closure criticality and/or are affected by the consequences of post-closure criticality. In addition, FEPs not previously considered are identified for further development.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

ADVANCES IN HEXAVALENT CHROMIUM REMOVAL AT HANFORD

At the Hanford Site, chromium was used as a corrosion inhibitor in the reactor cooling water and was introduced into the groundwater as a result of planned and unplanned discharges from reactors during plutonium production since 1944. Beginning in 1995, groundwater treatment methods were evaluated leading to the use of pump and treat facilities with ion exchange using Dowex 21 K, a regenerable strong base anion exchange resin. This required regeneration of the resin, which is currently performed offsite. Resin was installed in a 4 vessel train, with resin removal required from the lead vessel approximately once a month. In 2007, there were 8 trains (32 vessels) in operation. In 2008, DOE recognized that regulatory agreements would require significant expansion in the groundwater chromium treatment capacity. Previous experience from one of the DOE project managers led to identification of a possible alternative resin, and the contractor was requested to evaluate alternative resins for both cost and programmatic risk reductions. Testing was performed onsite in 2009 and 2010, using a variety of potential resins in two separate facilities with groundwater from specific remediation sites to demonstrate resin performance in the specific groundwater chemistry at each site. The testing demonstrated that a weak base anion single-use resin, ResinTech SIR-700, was effective at removing chromium, had a significantly higher capacity, could be disposed of efficiently on site, and would eliminate the complexities and programmatic risks from sampling, packaging, transportation and return of resin for regeneration. This resin was installed in Hanford's newest groundwater treatment facility, called 100-DX, which began operations in November, 2010, and used in a sister facility, 100-HX, which started up in September of 2011. This increased chromium treatment capacity to 25 trains (100 vessels). The resin is also being tested in existing facilities that utilize Dowex 21 K for conversion to the new resin. This paper will describe the results of the testing, performance in the facilities, continued optimization in the pump and treat facilities, and the estimated savings and non-tangible benefits of the conversion.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

The Off-Site Source Recovery Program Capabilities in Recovering Transuranic Sealed Sources - 17599

Over the past 16 years, the mission of the Off-Site Source Recovery Program (OSRP) has been recovering disused radioactive sealed sources in the interest of national security, public health and safety. This program, which is sponsored by the National Nuclear Security Administration's Office of Global Material Security's Office of Radiological Security, is managed at Los Alamos National Laboratory and Idaho National Laboratory and has recovered sources worldwide. To date, OSRP has recovered more than 38,000 sources from over 1,300 locations worldwide. Since its inception, OSRP has grown from recovering only sealed sources containing one transuranic (TRU) isotope, Pu-239, in the early days of the program, to recovering sealed sources containing several other TRU and non-TRU isotopes. Primarily, OSRP's mission includes Pu-238, Pu-239, Am-241, Cf-252, Cm-244, Sr-90, Cs-137, Co-60, Ir-192 and Ra-226 sources that are US-origin, defense related and not commercially disposable. Over the lifetime of this program the types of recoveries performed have expanded from a single site to multi-site round-up recoveries. An additional approach, referred to as self-ship recoveries, is also used in cases where only a few sources may be present at a licensee's site. In these situations, the licensee ships their source(s) to an OSRP staging facility for ultimate consolidation and disposal. OSRP recovers sources from many types of licensees, with the majority of sources being removed from industry, universities and hospitals. These include sealed sources used in industrial radiological devices like moisture-density gauges and thickness gauges, medical devices such as blood irradiators and radiotherapy devices and sealed sources from nuclear power plants, among others. The recovery and management of sealed sources is a complex process covering multidisciplinary areas, including source construction, characteristics of the radioactive material, source packaging and transportation per DOT regulations, safe and secure storage, and final disposition per waste acceptance criteria of repository sites. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2024

Nevada National Security Site - Area 5 Waste Received Under Permit SW532

Calendar Year 23 Asbestos Report for Permit SW532 Being resubmitted to address a comment received by NDEP, the weight and volume of two packages on one of the shipments was rounded differently than the quarterly report and we are revising the annual report to match.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Development of a Waste Management Organization for the Whiteshell Laboratories Closure Project - 18317

Whiteshell Laboratories (WL) has been generating and managing radioactive waste for over fifty years in support of a wide variety of research and operational applications. The current mission of the WL site, located in Manitoba, Canada, is to fully decommission the site to an approved end-state criterion and to disposition the waste in compliance with national and provincial regulations. To expedite the decommissioning schedule from over thirty years to nine years, which was a contractual commitment for the site operating company, Canadian Nuclear Laboratories (CNL), it became apparent that a centralized and optimized organization be established to coordinate all waste management activities and to fulfill the mission to process, condition, transport and disposition all radioactive wastes, which include Low-Level Waste (LLW), Intermediate-Level Waste (ILW) and High-Level Waste (HLW). The WL Waste Management Division was effectively established as a two-person operation in 2016 and has rapidly expanded to fulfill the mandate of managing nearly 30,000 cubic meters of LLW, ILW and HLW. This paper outlines the early lessons learned from managing important organizational changes without disrupting ongoing work practices; developing executive-level waste management strategies based on options studies and cost-benefit analysis comparisons; selection of storage vs. disposal disposition strategies; evaluation and selection of transportation methods and packages for LLW, ILW and HLW; and the need to develop an entire suite of programmatic documentation and procedures to support and instill a strong conduct of operations culture. The waste management strategy, which now focuses on transporting the waste to another province (Ontario) in a disposal ready form, represents the first large-scale radioactive waste transportation campaign in Canada that includes all three classes of waste. This paper provides additional insight on the overall waste management strategies for the WL Closure Project. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Decommissioning sodium dump tanks - 15547

The Nuclear Decommissioning Authority (NDA) is a non-departmental public body of the UK Government responsible for managing the effective and efficient clean-up of the UK's nuclear legacy. DSRL (Dounreay Site Restoration Limited, a wholly-owned subsidiary of the Cavendish Dounreay Partnership Ltd, a consortium of Cavendish Nuclear, CH2M HILL and URS) is the site licence company operating under contract to the NDA and funded by the NDA for the clean-up and demolition of the Dounreay Site, Britain's former centre of fast reactor research and development, near Thurso, Scotland, including decommissioning the Dounreay Prototype Fast Reactor (PFR). The plutonium metal fuel PFR was the second and last fast reactor to be built in the UK, with construction commencing in 1968. The reactor closed in 1994 and was de-fuelled, but the reactive sodium remaining in the cooling system required decommissioning, and approximately 12 tonnes of low level waste (LLW) sodium remained in the un-heated dump tanks. This paper highlights the approach used for decommissioning the dump tanks and their contents. The sodium was removed as a solid at ambient temperature, packaged, and staged for destruction at a later date. The tanks and pipework were then cleaned, size reduced and packaged for disposal as LLW. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

A methodology for optimal MSW management, with an application in the waste transportation of Attica Region, Greece

Highlights: • A two-step (strategic and detailed optimal planning) methodology is used for solving complex MSW management problems. • A software package is outlined, which can be used for generating detailed optimal plans. • Sensitivity analysis compares alternative scenarios that address objections and/or wishes of local communities. • A case study shows the application of the above procedure in practice and demonstrates the results and benefits obtained. - Abstract: The paper describes a software system capable of formulating alternative optimal Municipal Solid Wastes (MSWs) management plans, each of which meets a set of constraints that may reflect selected objections and/or wishes of local communities. The objective function to be minimized in each plan is the sum of the annualized capital investment and annual operating cost of all transportation, treatment and final disposal operations involved, taking into consideration the possible income from the sale of products and any other financial incentives or disincentives that may exist. For each plan formulated, the system generates several reports that define the plan, analyze its cost elements and yield an indicative profile of selected types of installations, as well as data files that facilitate the geographic representation of the optimal solution in maps through the use of GIS. A number of these reports compare the technical and economic data from all scenarios considered at the study area, municipality and installation level constituting in effect sensitivity analysis. The generation of alternative plans offers local authorities the opportunity of choice and the results of the sensitivity analysis allow them to choose wisely and with consensus. The paper presents also an application of this software system in the capital Region of Attica in Greece, for the purpose of developing an optimal waste transportation system in line with its approved waste management plan. The formulated plan was able to: (a) serve 113 Municipalities and Communities that generate nearly 2 million t/y of comingled MSW with distinctly different waste collection patterns, (b) take into consideration several existing waste transfer stations (WTS) and optimize their use within the overall plan, (c) select the most appropriate sites among the potentially suitable (new and in use) ones, (d) generate the optimal profile of each WTS proposed, and (e) perform sensitivity analysis so as to define the impact of selected sets of constraints (limitations in the availability of sites and in the capacity of their installations) on the design and cost of the ensuing optimal waste transfer system. The results show that optimal planning offers significant economic savings to municipalities, while reducing at the same time the present levels of traffic, fuel consumptions and air emissions in the congested Athens basin.

42 ENGINEERING↗
DOE OSTI2020

The TRANSCEND University Consortium: Theme 4: Nuclear Materials - 20431

The safe and secure management of Pu is a matter of international concern, with ∼250 t of separated Pu currently stockpiled worldwide. The UK's civil inventory of nuclear materials contains significant stocks of separated Pu from the reprocessing of Magnox and AGR spent fuels. The preferred option for the 138.5 tonnes of Pu is re-use as mixed oxide (MOx) fuel, although 5% is not suitable for re-use and is recommended for direct disposal. However, it will take more than 15 years to implement re-use, requiring that the Pu be kept in interim storage in its current state for that period, i.e. as PuO{sub 2} powder within inert steel storage cans at Sellafield. The focus of the work presented here is thus plutonium storage and the direct disposal of plutonium. The Research and Development needs of both are now pressing: in the case of storage due to it being the current default; in the case of immobilization and disposal because of a comparative lack of Research and Development on Pu conditioning and packaging due to policy uncertainty as to whether it would be disposed of in a Geologic Disposal Facility (GDF). Addressing these needs is complicated by Pu's high radioactivity, decay heat and radiotoxicity, criticality, nuclear safeguard requirements and, for some UK Pu contaminated materials targeted for disposal, poor inventory. Thus, there is also a critical requirement for underpinning research on Pu bearing materials in these two contexts. In response to these needs, the TRANSCEND Consortium (Transformative Science and Engineering for Nuclear Decommissioning, a multi-disciplinary collaboration of 11 universities and 8 key industry partners from across the UK's civil nuclear sector) is seeking to provide technical underpinning to ongoing option development for the UK's civil Pu stockpile. Whilst understanding the behaviour of plutonium during its re-use as MOx is beyond the scope of the TRANSCEND Consortium work plan, the main objectives of the work are: (1) For interim storage: to understand how the surface structure and properties of pristine and radiation damaged PuO{sub 2} change with time in the absence and presence of water; and (2) For immobilization and disposal: to understand the mechanisms of incorporation of Pu into ceramic and glass-ceramic waste-forms, as well as the effect on these of self-induced radiation damage. Each objective is being addressed through separate work packages, the details of which are discussed in this paper. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2012

NASA Advanced Explorations Systems: Concepts for Logistics to Living

The NASA Advanced Exploration Systems (AES) Logistics Reduction and Repurposing (LRR) project strives to enable a largely mission-independent cradle-to-grave-to-cradle approach to minimize logistics contributions to total mission architecture mass. The goals are to engineer logistics materials, common crew consumables, and container configurations to meet the following five basic goals: 1. Minimize intrinsic logistics mass and improve ground logistics flexibility. 2. Allow logistics components to be directly repurposed for on-orbit non-logistics functions (e.g., crew cabin outfitting) thereby indirectly reducing mass/volume. 3. Compact and process logistics that have not been directly repurposed to generate useful on-orbit components and/or compounds (e.g., radiation shielding, propellant, other usable chemical constituents). 4. Enable long-term stable storage and disposal of logistics end products that cannot be reused or repurposed (e.g., compaction for volume reduction, odor control, and maintenance of crew cabin hygienic conditions). 5. Allow vehicles in different mission phases to share logistics resources. This paper addresses the work being done to meet the second goal, the direct repurposing of logistics components to meet other on-orbit needs, through a strategy termed Logistics to Living (L2L). L2L has several areas but can be defined as repurposing or converting logistical items (bags, containers, foam, components, etc.) into useful crew items or life support augmentation on-orbit after they have provided their primary logistics function. The intent is that by repurposing items, dedicated crew items do not have to be launched and overall launch mass is decreased. For non-LEO missions, the vehicle interior volume will be relatively fixed so L2L will enable this volume to be used more effectively through reuse and rearrangement of logistical components. Past work in the area of L2L has already conceptually developed several potential technologies [Howe, Howard 2010]. Several of the L2L concepts that have shown the most potential in the past are based on NASA cargo transfer bags (CTBs) or their equivalents which are currently used to transfer cargo to and from the ISS. A high percentage of all logistics supplies are packaging mass and for a 6-month mission a crew of four might need over 100 CTBs. These CTBs are used for on-orbit transfer and storage but eventually becomes waste after use since down mass is very limited. The work being done in L2L also considering innovative interior habitat construction that integrate the CTBs into the walls of future habitats. The direct integration could provide multiple functions: launch packaging, stowage, radiation protection, water processing, life support augmentation, as well as structure. Reuse of these CTBs would reduce the amount of waste generated and also significantly reduce future up mass requirements for exploration missions. Also discussed here is the L2L water wall , an innovative reuse of an unfolded CTB as a passive water treatment system utilizing forward osmosis. The bags have been modified to have an inner membrane liner that allows them to purify wastewater. They may also provide a structural water-wall element that can be used to provide radiation protection and as a structural divider. Integration of the components into vehicle/habitat architecture and consideration of operations concepts and human factors will be discussed. In the future these bags could be designed to treat wastewater, concentrated brines, and solid wastes, and to dewater solid wastes and produce a bio-stabilized construction element. This paper will describe the follow-on work done in design, fabrication and demonstrations of various L2L concepts, including advanced CTBs for reuse/repurposing, internal outfitting studies and the CTB-based forward osmosis water wall.

Shull, Sarah A.↗
DOE OSTI2012

Integration of the informal sector into municipal solid waste management in the Philippines - What does it need?

The integration of the informal sector into municipal solid waste management is a challenge many developing countries face. In Iloilo City, Philippines around 220 tons of municipal solid waste are collected every day and disposed at a 10 ha large dumpsite. In order to improve the local waste management system the Local Government decided to develop a new Waste Management Center with integrated landfill. However, the proposed area is adjacent to the presently used dumpsite where more than 300 waste pickers dwell and depend on waste picking as their source of livelihood. The Local Government recognized the hidden threat imposed by the waste picker's presence for this development project and proposed various measures to integrate the informal sector into the municipal solid waste management (MSWM) program. As a key intervention a Waste Workers Association, called USWAG Calahunan Livelihood Association Inc. (UCLA) was initiated and registered as a formal business enterprise in May 2009. Up to date, UCLA counts 240 members who commit to follow certain rules and to work within a team that jointly recovers wasted materials. As a cooperative they are empowered to explore new livelihood options such as the recovery of Alternative Fuels for commercial (cement industry) and household use, production of compost and making of handicrafts out of used packages. These activities do not only provide alternative livelihood for them but also lessen the generation of leachate and Greenhouse Gases (GHG) emissions from waste disposal, whereby the life time of the proposed new sanitary landfill can be extended likewise.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2023

IACMI Project 4.7: Pultruded Textile Carbon Fiber for Spar Caps (Final Report)

The primary objective of this project was to demonstrate the potential to significantly reduce the cost of wind turbine blades with carbon fiber reinforced polymer (CFRP) structure. Applicability of textile carbon fibers (TCF) were evaluated for use in pultruded spar cap (SC) elements as a path to cost reduction for utility scale wind turbine blades. In earlier work for the Department of Energy (DOE) Wind Energy Technologies Office (WETO), a collaboration of Sandia National Laboratory (SNL), Oak Ridge National Laboratory (ORNL), and Montana State University has demonstrated potential for pultruded TCF to compete with infused fiberglass and commercially available carbon fiber pultruded sections for spar cap construction. In the design cases evaluated, the TCF sections fared well when compared on cost per unit composite stiffness and cost per unit composite compressive strength for those designs [1]. Both stiffness and compressive strength tend to be key factors in the design of blade composite Spar Cap which carry the bulk of the blade structural loads in bending. Spar Cap design tends to distribute largely symmetric tensile and compressive stresses to opposite sides of the spar structure, but since carbon fiber composite compressive strength is typically 20-50% lower than tensile strength, the compressive loading reaches failure levels well before the tensile loading. Stiffness is critical in containing the large tip deflection in high wind loading situations. However, materials and process development were very limited in the earlier study and the work in this project was expanded to make the comparative information more representative of what will be required in order to make further inroads towards implementation. Similar to that study, this project team confirmed that the primary materials of interest for pultruded spar cap elements should be thermoset (TS) resins reinforced by carbon fibers, utilizing as high a percentage of TCF as practical to benchmark cost and performance against commercial carbon fibers. To make the closest comparison possible and eliminate specific test article size, resin selection, and equipment/operational nuance effects, the team planned to pultrude sections with 100% commercially available carbon fiber (Panex 35 carbon fiber from Zoltek) as well as samples utilizing high fractions of TCF. The resin system chosen was based on formulations recommended by large wind industry supplier Hexion and consisted of Hexion resin RSL-4597, curing agent CCA-138, and internal mold release additive 117, along with common kaolin filler ASP400P from BASF. As commonly deployed in spar cap configurations, the team had a mold built to pultrude a rectangular spar cap element of 100mm width and 3mm thickness. The extremely limited number of samples produced for the earlier study were produced with a “generic” epoxy utilized for a variety of applications by the pultruder contracted to produce test articles for demonstration purposes. More importantly, those samples were produced at a fiber fraction only slightly over 50%. Based on feedback from our industrial advisory team for that project and strongly recommended by this project team, the consensus is that it is highly desirable to obtain fiber fractions of 65-68% for significant penetration in wind blade spar cap. Although this requirement has yet to be exhaustively confirmed in readily available information, this was established as a project goal and informally decided we needed to exceed 60% fiber fraction to gain serious industry consideration. Previous TCF pultrusion trials have been challenged by the lack of robust TCF packages, resulting in non-uniform tension across and between tows, as well as excess labor and waste for removal of interleaved paper. The non-uniform tension and associated intermingling of tows in textile acrylic fiber tows and associated difficulties created from broken filaments in carbon fiber conversion inhibit the ordered packing necessary to enhance fiber fraction elevation. (These “cross-overs” are not considered undesirable for textile applications and there is some sense that they might be advantageous for those applications). In addition to work that is ongoing at the acrylic fiber manufacturers to improve their formats, The Institute of Advanced Composites Manufacturing Innovation (IACMI) Project 6.12 (report PA16-0349-6.12-01) [2] has developed and demonstrated a more robust packaging and creeling approach that at least partially addresses these issues, thus improving control of the TCF feed into the pultrusion unit. It was hoped that these and other improvements currently being implemented would allow us to achieve fiber fractions at least approaching these fiber fraction targets. During this project, sections utilizing 100% commercially available carbon fiber reinforcement were produced as a baseline, as well as sections reinforced with about 94% TCF and the balance being commercially available fiber for comparison. The most important finding was that similar to results reported in the earlier WETO-funded project and results from tests of TCF reported at IACMI meetings, this work demonstrated that sections pultruded with TCF in an epoxy resin frequently utilized in actual spar cap production had stiffness and compressive strengths largely comparable to similar sections pultruded with a commercially available carbon fiber also frequently utilized in the wind industry. Although the amount of that data is limited, some of the tensile strength results were actually closer than would have been expected based on fiber strength results provided by the TCF and commercial fiber producers. The actual test data are reported and discussed in detail in Section 5. The pultruded sections dominated by TCF reinforcement were approximately 8-10% lower in fiber fraction than for the sections produced using commercial fiber alone, making direct comparison difficult. The COVID-19 project has provided significant insight into the current state-of-the-art with various TCF product forms. The data obtained in this project will guide the planned improvements at the precursor level, especially in attaining uniform tensioning and payout to facilitate enhanced fiber fractions and overall processability of the TCF composites. The project team is providing guidance to stakeholders concerning the attributes, needs, and potential demand for TCF in wind blade spar caps. Results achieved in this project are consistent with findings in the related work cited [1] and support this guidance and the high potential for this product type. TCF precursor-producing partners continue to express interest in enhancing their product forms and the team looks forward to working with these improved materials as they become available.

42 ENGINEERING↗
DOE OSTI2013

Production and characterization of a composite insulation material from waste polyethylene teraphtalates

The pollution of polyethylene teraphtalate (PET) is in huge amounts due to the most widely usage as a packaging material in several industries. Regional pumice has several desirable characteristics such as porous structure, low-cost and light-weight. Considering the requirements approved by the Ministry of Public Works on isolation, composite insulation material consisting of PET and pumice was studied. Sheets of composites differing both in particle size of pumice and composition of polymer were produced by hot-molding technique. Characterization of new composite material was achieved by measuring its weight, density, flammability, endurance against both to common acids and bases, and to a force applied, heat insulation and water adsorption capacity. The results of the study showed that produced composite material is an alternative building material due to its desirable characteristics; low weight, capability of low heat conduction.

36 MATERIALS SCIENCE↗