Mindess Developments in the Formulation and Reinforcement of Concrete

1. Auflage 2014
ISBN: 978-1-84569-468-5
Verlag: Elsevier Science & Techn.
Format: EPUB
Kopierschutz: 6 - ePub Watermark

E-Book, Englisch, 308 Seiten

Reihe: Woodhead Publishing Series in Civil and Structural Engineering

Concrete is the most widely-used construction material in the world. This important book summarises the wealth of recent research on improving qualities such as durability and sustainability as well as the emergence of a new generation of specialist concretes for particular applications.A number of chapters discuss new types of concrete such as autoclaved aerated concrete, high-strength concrete, sprayed concrete, fibre-reinforced concrete, lightweight concrete, self-compacting concrete, foamed and polymer concrete, together with their characteristics and applications. Other chapters review the development of concrete especially suited for particular conditions such as radiation protection, hot weather and underwater conditions, as well as the increasingly important area of recycling.With its distinguished editor and international team of contributors, Developments in the formulation and reinforcement of concrete is a standard reference for civil and structural engineers. - Summarises a wealth of recent research on improving qualities such as sustainability and durability - Discusses new concrete types together with their characteristics and applications - Reviews the development of concrete especially suited to particular conditions such as hot weather and under water

Autoren/Hrsg.

Mindess, Sidney

Leseproben

2 High-density and radiation-shielding concrete
P. Lessing    Idaho National Laboratory, USA 2.1 Introduction
2.1.1 Definition of high-density concrete
High-density concrete is produced using special heavy aggregates and can have a density of up to 400 lb/ft3. Why would anyone be interested in a heavyweight concrete when low-density concrete has so many beneficial construction and insulation applications? The answer is that heavyweight concrete is primarily useful for nuclear radiation shielding but can also be useful for counterweights and blast shielding (applications where physical space is limited). Gamma-rays and X-rays can be shielded by a mass of material containing heavy atoms. To shield against neutrons it is necessary to have a mass of material that contains atoms that can both “thermalize” and capture neutrons. The hydrogen atom in water that is chemically bonded in concrete “thermalizes” the fast neutrons which may then be captured by other atoms such as boron which have high neutron-capture cross-sections. Reference books are available that cover aggregate characteristics, mix proportioning, and standards for conventional high-density concrete.1,2 2.1.2 Characteristics of various heavy aggregates
Historically inexpensive high-density mineral aggregates such as barite, ferro-phosphorus, goethite, hematite, ilemnite, limonite, magnetite, and steel punching and steel shot have been used to produce high-density concrete. Table 2.1 shows typical densities of heavy aggregates and the concrete manufactured using these aggregates. Table 2.1 Physical properties of traditional heavyweight aggregates and concrete Geothite (hydrous iron ore) 10–11 3.4–3.7 130–140 180–200 Barite (barium sulfate) 0 4.0–4.6 145–160 210–230 Hematite (iron ores) – 4.9–45.3 180–200 240–260 Magnetite (iron mineral) – 4.2–5.2 150–190 210–260 Ferro-phosphorus (slag) 0 5.8–6.8 200–260 255–330 Steel punchings or shot 0 6.2–7.8 230–290 290–380 Boron-containing additions such as colemanite, boron frits, and borocalcite have been used to improve the neutron-shielding properties of heavy concrete. However, they may negatively affect the setting and strength of the concrete. 2.1.3 Requirements for radiation-shielding concrete
High level functional requirements for spent nuclear fuel storage applications were gathered into a report by Haelsig.3 Haelsig tabulates the requirements that he determined to be applicable to the conceptual design of a multi-purpose cask (MPC) and allocates them to MPC subsystems. In several appendices, Haelsig lists the design criteria of several vendor supplied concrete casks that were then licensed by the Nuclear Regulatory Commission (NRC). The design criteria/requirements vary by manufacturer, but are in the general categories of: service life, criticality safety limit, surface contact dose, dose for storage, storage facility dose limits, ambient environment, fuel cladding temperature limits, maximum decay heat power, concrete temperature limits, canister internal pressure limits and leak tightness, seismic ground accelerations, tornado loads, flooding loads basis, snow and ice loads, cask drops and fuel impact acceleration, and overall shipping package width on transport vehicle. Obviously, the mechanical and physical properties of a heavy concrete (e.g., attenuation of radiation, density, strength, thermal conductivity, etc.) have an impact on how a cask design meets most of these functional requirements. For instance, the Haelsig report calculates how the wall thickness will be greatly reduced using a heavy versus ordinary concrete while still meeting the radiation criteria.4 However, no physical properties are specified for the concrete (other than in-service temperature requirements). The physical properties of a specific grade of concrete need to be coupled with a specific cask design in order to meet the overall functional requirements. Table 2.2 shows the important aspect of temperature requirements for concrete spent fuel storage casks listed by various manufacturers. Pacific Nuclear and Sierra Nuclear appear to interpret NRC guidance somewhat differently. Pacific Nuclear’s criteria distinguish between duration and location (short versus long term, surface versus bulk, and local). Examining Table 2.2, it appears that the long-term maximum exposure (local) temperature for concrete in spent nuclear fuel storage casks should be less than 149 °C (300 °F). This is fairly consistent with a performance test sponsored by the Electric Power Research Institute (EPRI).8 Table 2.2 Maximum temperatures of concrete in storage cask Depleted Uranium ACI-349 66 °C 93–149 °C 177–343 °C Concrete Container Feasibility Study5 Appendix A and NRC Guidance (150 °F) (200–300 °F) (350–650 °F) Pacific Nuclear ACI-349 = 66 °C = 93 °C (= 200 °F) 177(surface)– “Nuhoms” System Appendix A (= 150 °F) 343 °C (local) (350–650 °F) Sierra Nuclear ACI-349 66 °C 149 °C (local) “VSC” System6 Appendix A and NRC Guidance (150 °F) (300 °F) Babcock & Wilcox Fuel Company7 121 °C (250 °F) A relevant standard from the American Cement Institute (ACI) standard 349 Appendix A.4 states: A.4.1- The following temperature limitations are for normal operation or any other long-term period. The temperatures shall not exceed 150 °F except for local areas, such as around penetrations, which are allowed to have increased temperatures not to exceed 200 °F …” ACI-349 Appendix A.4 was obviously written for conventional concrete using rock aggregate such as quartzite and not for synthetic aggregate. Therefore, to qualify under ACI-349, long-term exposures should be kept under 200 °F (93.3 °C). Using this standard, concrete fabricated with synthetic aggregate would qualify if the strength did not deteriorate at temperatures from 90 °C to 125 °C. 2.2 Applications/case studies
In the United States, depleted uranium (uranium having 235U content less than natural uranium’s 0.711 wt%) has been generated as tails from uranium enrichment and spent fuel reprocessing. This comprises approximately 500,000 metric tons (uranium content). About 470,000 metric tons are stored as pure DUF6 in steel cylinders at US gaseous diffusion enrichment sites. There is approximately 225,000 metric tons of elemental fluorine associated with this stored DUF6. Intact cylinders normally contain DUF6 with purity exceeding 99.9%. In addition, the UDOE-EM currently owns about 19,500 metric tons of elemental uranium (MTU) in the form of high purity DUO3 resulting from historical weapons production programs at US defense complexes. In 2002, the DOE awarded Uranium Disposition Services, LLC (UDS) a contract to design, build and operate two DUF6 conversion facilities at Paducah, Kentucky and Portsmouth, Ohio. The facilities were designed to convert DUF6 into uranium oxide for disposal (or an alternate use) and aqueous hydrogen fluoride which is to be sold. The facilities were expected to be in full operation by June 2008.9 When the conversion facilities are operational, large quantities of high purity depleted uranium oxide powder will be available. This material could be used as a feedstock to manufacture aggregate suitable for inclusion into radiation shielding heavy concrete. 2.3 The case of DUAGG® and DUCRETE®
In the mid-1990s the Idaho National Laboratory (INL) developed new methods to produce high-density aggregate (synthetic rock) primarily consisting of depleted uranium oxide.10 The objective was to develop a low-cost method whereby depleted uranium oxide powder (UO2, U3O8, or UO3) could be processed to produce high-density aggregate pieces (DUAGG) having physical properties suitable for disposal in low-level radioactive disposal facilities or for use as a component of high-density concrete used as shielding for radioactive...