Survey of Generation IV Nuclear Reactors.
\r\n\t a multi-pronged approach. The pervasive computing paradigm is at a crossroads where never before computing
\r\n\t has been so much embedded within the user. Recent developments in sensor technologies, wireless protocols
\r\n\tintegration, and AI have empowered the citizen towards a smart citizen with a high degree of autonomy and varying
\r\n\tcomputing capabilities from one context to another.
\r\n\t
\r\n\tMoreover, software engineering has evolved too to allow lightweight programming and full-stack coding of those sensors. The network itself is today viewed as a programming platform, thus wearable devices are no more stand-alone and do not operate in a vacuum. This book aims at attracting authors from academia, the industry, research institutions, public and private agencies to provide the findings of their recent achievements in the field, but also visionaries who foresee the future of wearable technologies in the coming decades.
Fifty years ago, on June 26, 1954, in the town of Obninsk, near Moscow in the former USSR, the first nuclear power plant was connected to an electricity grid to provide power. This was the world\'s first nuclear power plant to generate electricity for a power grid, and produced around 5 MWe [1]. This first nuclear reactor was built twelve years after the occurrence of the first controlled fission reaction on December 2, 1942, at the Manhattan Engineering District, in Chicago, Illinois, US. In 1955 the USS Nautilus, the first nuclear propelled submarine, equipped with a pressurized water reactor (PWR), was launched. The race for nuclear technology spanned several countries and soon commercial reactors, called first generation nuclear reactors, were built in the US (Shippingport, a 60 MWe PWR, operated 1957-1982, Dresden, a boiling water reactor, BWR, operated 1960-1978, and Fermi I, a fast breeder reactor, operated 1957-1972) and the United Kingdom (Magnox, a pressurized, carbon dioxide cooled, graphite-moderated reactor using natural uranium).
\nA few years after the projects had developed many nuclear safety concepts were extended and then implemented in second-generation nuclear systems, consisting of reactors currently in operation, as the PWR, CANDU (Canadian Deuterium Uranium Reactor), BWR (Boiling Water Reactor), GCR (Gas-Cooled Reactor), and VVER (Water - Water Power Reactor), the latter developed by the former Soviet Union. At this time, other concepts were studied in parallel, such as liquid metal cooled reactors and the reactors with thorium and uranium molten salt, which did not propagate commercially and/or remained in experimental countertops. Operating or decommissioned power reactor designs can be found in [2].
\nWith operating experience gained in recent decades, digital instrumentation development and lessons learned from the accidents at TMI (Three Mile Island), Chernobyl and recently Fukushima, Generation III and III+ reactor designs have incorporated improvements in thermal efficiency and included passive system safety and maintenance costs and capital reduction. There are several designs of these so called advanced reactors and some are being built in the U.S. and China.
\nIn 2001, nine countries (Argentina, Republic of Korea, Brazil, Canada, Republic of South Africa, United Kingdom, France, United States and Japan) signed the founding document of Generation IV International Forum (GIF) in order to develop nuclear systems that can fulfill the increasing world electric power needs with high safety, economics, sustainability and proliferation resistance levels [3]. The Russian Federation, People´s Republic of China, Switzerland and Euratom have joined this group. Since then, GIF has selected the six most promising reactor system designs to be developed until 2030 and has created research groups on materials, fuel and fuel cycle, conceptual design, safety, thermal-hydraulics, computational methods validation and others in order to develop the necessary technology on an international cooperation basis.
\nAccording to IAEA [4], an accessible, affordable and sustainable energy source is fundamental to the development of modern society. Current scenarios predict a global demand for primary energy 1.5-3 times higher in 2050 as compared to today, and a 200% relative increase in the demand for electricity. Nuclear power is an important source that should be considered, because it is stable power on a large scale, with virtually no greenhouse gas emissions and low environmental impact as compared to fossil fuels, and can produce heat for chemical processes in industry and for hydrogen generation. In this context, GIF is seeking to develop more economical, sustainable and safe nuclear reactors, from their fuel cycles to decommissioning and waste treatment, and thus meet the world’s energy needs.
\nThis chapter is organized as follows. Section 2 discusses the goals for Generation IV reactor systems, section 3 discusses the current six Generation IV reactor system design description and research on the subject. Section 4 focuses on the first reports concerning reactor safety and risks [5], particularly the Integrated Safety Assessment Methodology (ISAM). Section 5 addresses economic aspects of this new reactor generation [6]. Section 6 concerns the discussion on proliferation resistance and physical protection [7]. Section 7 encompasses a set of general conclusions on the subject, addressing mainly the relevance of these nuclear system concepts.
\nBefore selecting the reactors that will be part of the Next Generation Nuclear Plant (NGNP), the founding countries of the Generation IV International Forum selected eight key objectives able to make these reactor designs vital in the near future. These eight goals based on concepts of sustainability, economic competitiveness, safety, physical protection and nuclear proliferation resistance are described below [3]:
\nSustainability-1 – NGNP will provide sustainable energy generation that meets clean air objectives and provides long-term system availability and effective fuel utilization for worldwide energy production;
Sustainability-2 – NGNP will minimize and manage their nuclear waste and notably reduce the long-term administrative burden, thereby improving protection for the public health and the environment.
Economics-1 – NGNP will have a clear life-cycle cost advantage over other energy sources.
Economics-2 – NGNP will have a level of financial risk comparable to other energy projects.
Safety and Reliability-1 – NGNP operations will excel in safety and reliability.
Safety and Reliability-2 – NGNP will have a very low frequency and degree of reactor core damage.
Safety and Reliability-3 – NGNP will eliminate the need for offsite emergency response.
Proliferation resistance and Physical Protection – NGNP will increase the assurance that they are very unattractive and the least desirable route for diversion or theft of weapons-usable materials, and provide increased physical protection against acts of terrorism.
With these concepts in mind, projects already in operation, in test or just conceptual have been analyzed and six promising nuclear reactors have been selected to be designed and built in this endeavor.
\nBased on GIF goals, six reactor designs have been selected to be developed and constructed by 2030. Such reactors must meet the safety and security, sustainability, economics and proliferation resistance criteria defined as essential for this new generation. That is, projects must: consider the entire fuel cycle to obtain a higher fuel burn-up and consequently less actinides in the final waste, reducing their lifetime in final repositories; increase the thermal efficiency with combined cycles and cogeneration; aim at the production of hydrogen, be intrinsically safe, with negative reactivity coefficients and achieve competitive costs since their construction.
\nGIF has divided the involved countries in working groups formed by laboratories, universities, and government agencies, according to the experience and interest of each to develop the projects of the following reactors: Very-high-temperature reactor (VHTR), Gas-cooled fast reactor (GFR), Supercritical-water-cooled reactor (SCWR), Sodium-cooled fast reactor (SFR), Lead-cooled fast reactor (LFR) and Molten salt reactor (MSR). For each project specific groups of materials, computational methods, fuel and fuel cycle, thermal-hydraulics, safety and operation and others have been created. The first four reactor concepts are completely defined and the remaining two are in progress.
\nA major challenge for almost every project is the development of new materials for the reactor primary systems that can tolerate temperatures up to 1000 °C without reducing safety margins. Another point is the need to develop and validate computer codes both in neutronics and thermal-hydraulics projects with little or no available operational experience. These issues and the obtained solutions will be discussed next.
\nThe VHTR is one of the most promising projects of Generation IV reactors, given the experience in gas-cooled reactors that many countries developed in recent decades. The first reactors of this type were designed by the UK and France for the production of plutonium [8], were graphite-moderated and used CO2 as a coolant, which limited the maximum temperature at 640 oC, when chemical reactions occur between gas and moderator. Later, the use of helium (although more costly) was justified by having better heat transfer properties, and permit increasing the core outlet temperature.
\nThe current GIF design consists of a helium-cooled reactor with graphite as a moderator, TRISO fuel and coolant outlet temperature above 900 oC. There are two core concepts: the prismatic block-type and the pebble bed-type. The first type follows the line of the High Temperature Engineering Test Reactor (HTTR) developed and built by Japan initially with coolant exit temperature of 850 oC and then 950 oC in April 2004 [9]. The second is the result of the German program, which was later imported by the People\'s Republic of China and developed in the Republic of South Africa as the Pebble Bed Modular Reactor (PBMR). Both designs use TRISO fuel (see Figure 1), which consists of a spherical micro kernel of oxide or oxycarbide fuel and coating layers of porous pyrolytic carbon (buffer), inner dense pyrolytic carbon (IPyC), silicon carbide (SiC) and dense outer pyrolytic carbon (OPyC).
\nTRISO fuel for prismatic block-type and pebble bed-type cores [10,11].
The goal of achieving an exit temperature of 1000oC coolant will allow the VHTR to generate electricity with high efficiency and provide heat for hydrogen production and for refineries, petrochemical and metallurgical industries. For this purpose, research is being done to evaluate the use of other materials, such as uranium-oxicarbide UCO and ZrC, which increase the capacity of TRISO fuel burn-up, reduce the permeability of fission products and increase resistance to high temperatures in case of an accident (above 1600 °C) [12]. Although a once-through uranium fuel cycle is planned reactors have potential for this deep-burn of plutonium and minor actinides, as well as the use of thorium based fuel [13].
\nThe experimental reactor HTTR in Japan and HTR-10 in China support the development of the VHTR, in particular providing important information on safety and operational characteristics. The research group on computational methods validation and benchmarking uses these reactors to validate computer codes in the areas of thermal hydraulics, thermal mechanics, core physics and chemical transport.
\nAnother project selected by GIF of a gas-cooled reactor is the GFR, the fast-neutron-spectrum, helium-cooled and closed fuel cycle reactor. The selection of this reactor was based on its excellent potential for long-term sustainability, in terms of the use of uranium and of minimizing waste through reprocessing and fission of long-lived actinides [14,15]. In terms of non-proliferation, the objective of high burn-up together with actinide recycling results in spent fuel that is unattractive for handling. The use of gas as coolant means achieving high temperatures, so that this design also has the purpose of providing process heat and enabling hydrogen production.
\nOne of the proposed designs is a 1200 MWe reactor, operating at 7 MPa, coolant exit temperature of 850oC, an indirect combined cycle with He-N2 gas mixture for intermediate gas cycle and natural convection as passive safety system [14]. At least two fuel concepts that meet the proposed design are being studied: a ceramic plate-type fuel assembly and a ceramic pin-type fuel assembly. In this latter, the fuel assembly is based on a hexagonal lattice of fuel-pins and the materials used are uranium and plutonium carbide as fuel, and silicon carbide as cladding [16].
\nAn experimental reactor called ALLEGRO (Fig. 2) with 80 MWth will be the first built GFR with the objective of demonstrating the feasibility and qualifying technologies for fuel, fuel assembly and new safety systems. It is important to notice the interface between GFR and VHTR reactors, which use helium as a coolant and power conversion technology using gas turbines and cogeneration, as many efforts on component and materials development are being held together.
\nUpdated layout of ALLEGRO featuring two main heat exchange loops [15].
The SCWR concept emerged at the University of Tokyo in 1989 and became a global concern after being selected by the Generation IV International Forum in 2002 [17]. The SCWR is a high-temperature, high-pressure, water-cooled reactor that operates above the thermodynamic critical point of water (374 oC, 22.1 MPa). This last feature eliminates coolant boiling, since both liquid and gaseous states can coexist. Thus, according to [18], the need for recirculation and jet pumps, pressurizer, steam generators, and steam separators and dryers in current LWRs is eliminated. Still, according to [18] the main mission of this reactor design is the generation of low-cost electricity.
\nA more advanced project is the Japanese 1620 MWe SCWR consisting of a pressure-vessel type, once-through reactor and a direct Rankine cycle system. Figure 3 is a simplified system where steam leaves the pressure vessel at 560 oC and 25 MPa, passes through moisture separator valves and control valves, high and low pressure turbines connected to the generator, condenser, low and high pressure pumps, a deaerator (to remove dissolved gases) to return to the pressure vessel. Because of supercritical water condition, no phase change occurs, which means that the coolant continuously passes from the liquid state to the gaseous state.
\nThe Japanese Supercritical Water Reactor (JSCWR) [17].
Another concept proposed by Canada is generically called CANDU-SCWR [19], it is a pressured tube type reactor with fuel channels separating the light water coolant from the heavy water moderator. The supercritical steam is led directly to the high-pressure turbine, eliminating the need of steam generators as the JSCWR.
\nThe Sodium-cooled Fast Reactor is among the six candidate designs selected for their potential to meet Generation IV technology goals. It is a fast-spectrum, sodium-cooled reactor and closed fuel cycle for efficient management of actinides and conversion of fertile uranium Three options are under consideration (Figure 4): a large size loop type sodium-cooled reactor with uranium-plutonium oxide fuel (1500 MWe), a medium size pool-type system with 600 MWe and a small size modular type with plutonium-minor actinide-zirconium metal alloy fuel [20].
\nLoop, modular and pool configuration [20,21]
Heat capacity and high thermal conductivity of liquid-metal coolants provides large thermal inertia against system heating during loss-of-flow accidents. Furthermore, due the non-corrosive characteristic of sodium coolant, the reactor core and primary system components do not degrade even over very long residence times in the reactor, so that maintenance requirements are minimized. However sodium reacts chemically with air and water and requires a great coolant system seal.
\nThe application of lead coolant in nuclear systems began in the 1970s in the ancient Soviet Union with systems cooled by Lead-Bismuth Eutectic (LBE), which were developed for nuclear-powered submarines [22]. Two of the currently proposed designs for international cooperation are the Small Secure Transportable Autonomous Reactor (SSTAR, Fig. 5) and the European Lead-cooled System (ELSY). Both concepts are lead-cooled, with passive decay heat removal and nitride fuels. A closed fuel cycle is expected for the efficient conversion of fertile uranium and actinide management for this reactor.
\nSmall Secure Transportable Autonomous Reactor [22].
Lead flows in the reactor core cooling it by natural circulation and passes through Pb-to-CO2 heat exchangers located between the vessel and the cylindrical shroud. It is noteworthy that lead has been chosen as coolant rather than LBE to drastically reduce the amount of alpha-emitting 210Po isotope produced [22]. Many studies on the choice of fuel and fuel cycle have been proposed since 2004 when an international cooperation through GIF began. Features of this reactor, like enabling fissile self-sufficiency, autonomous load following, operation simplicity, reliability, transportability, as well as a high degree of passive safety, make it a unique proliferation resistant, safe and economical design.
\nInvestigation of molten-salt reactors started in the late 1940s as part of the United States\' program to develop a nuclear powered airplane [23]. Today, research on this reactor has been resumed due to its inclusion as one of the six Generation IV reactor types. MSRs have been initially considered as thermal-neutronic-spectrum graphited-moderated concepts, but since 2005 R&D has focused on the development of fast-spectrum MSR concepts. The Molten Salt Fast Reactor core is a cylinder of the same diameter and height, where nuclear reactions occur in the fuel salt. Two options have been considered for the fuel cycle: 233U-started MSFR and TRU(fuelled with transuranic elements)-started MSFR [24]. The salt processing scheme relies on both on-line and batch processes (Fig. 6) to satisfy the constraints for a smooth reactor operation while minimizing losses to waste stream [25]. A simplified schematic of the online processing can be seen in Figure 6, where fuel fission gas extraction and actinide separation occur.
\nTMSR (MSFR) reference fuel salt processing [25].
A survey of the main features of the reactors discussed here is displayed in Table 1. It is important to remember that these projects are in development and there are different views in the same reactor line. Some reactors are already testing their designs with full construction and will be essential for future validation methodologies. If these six reactors achieve their ultimate goals, they will not only bring more advanced technology, but new concepts of safer, sustainable and economical nuclear reactors. Some of these aspects will be discussed in more detail in the following sections.
\n\n | \n GFR\n | \n\n SCWR\n | \n\n MSR\n | \n\n SFR\n | \n\n VHTR\n | \n\n LFR\n | \n
\n Neutrons energy\n | \nFast | \nThermal | \nThermal or fast | \nFast | \nThermal | \nFast | \n
\n Power (MWe)\n | \n1200 | \nUp to 1500 | \n1000 | \n50 - 1500 | \n~200 | \n19.8 - 600 | \n
\n Fuel type\n | \nPlate or ceramic | \nUO2 pellet | \nMolten salt | \nMOX or Metal alloy | \nTRISO | \nMOX or Nitrides | \n
\n Coolant\n | \nHelium | \nLight water | \nMolten salt | \nSodium | \nHelium | \nLead | \n
\n Moderador\n | \n- | \nLight water | \n- | \n- | \nGraphite | \n- | \n
\n Outlet temperature\n | \n850oC | \n625oC | \n1000oC | \n~500oC | \n1000oC | \n~500oC | \n
\n Pressure * (MPa)\n | \n7 | \n25 | \n~0.1 | \n~0.1 | \n5 - 7 | \n~0.1 | \n
\n Fuel Cycle\n | \nClosed | \nonce through | \nClosed | \nClosed | \nonce through | \nClosed | \n
Survey of Generation IV Nuclear Reactors.
* Primary system pressure.
Since the very beginning, Safety and Reliability issues were extensively considered, representing three of the eight goals of Generation IV Nuclear Energy Systems [26]. Safety, together with waste management and nuclear proliferation risks, remains as one of the key problems of nuclear energy. Consequently, a new generation of Nuclear Power Plants (NPPs) has to face and solve these problems convincingly.
\nIn spite of past nuclear accidents, including the most recent at Fukushima Daiichi, NPPs have an excellent safety record and still remain as a safe and high-power energy production technology without greenhouse gases emissions. Unfortunately, public acceptance continues to be an important impediment.
\nIn this context, the Generation IV International Forum (GIF) safety objectives have been oriented to substantially upgrade safety and enhance public confidence in NPPs, by means of an increasing use of inherent safety features, a major reduction of core damage frequency and by eliminating the need for offsite emergency response in case of accidents [26]. It was immediately recognized the necessity of a standard methodology for Generation IV safety assessments. The methodology would allow a uniform safety evaluation of different NPP concepts with respect to Generation IV safety goals [26].
\nDuring the past 60 years safety assessment has evolved, from the initial deterministic analysis through conservative assumptions and calculations, to an increasing application of a best-estimated deterministic approach, in conjunction with probabilistic methods [27, 28]. Best-estimated and probabilistic assessments identify potential accidents scenarios that could be important contributors to risk [28]. As a consequence, Probabilistic Safety Analysis (PSA) is having a more prominent and fundamental role in the design process and licensing analysis, as part of an integrated approach, risk-informed, which also includes Deterministic Analysis and Defense in Depth Philosophy [29]. This new safety assessment conception, integrating probabilistic and deterministic analysis with an extensive application of the Defense in Depth principle, constitutes the basis for the Generation IV safety assessment methodology.
\nIn 2004, the Idaho National Engineering and Environmental Laboratory proposed a Research and Development Program Plan, identifying the R&D needs for the next generation nuclear plant (NGNP) design methods [30]. Later, in 2008, the Idaho National Laboratory prepared a report for the US Department of Energy with a R&D Technical Program Plan for the NGNP methods [31]. Both reports were focused on the development of tools to assess the neutronic and thermal-hydraulic behavior of the Very High Temperature Reactor (VHTR) systems, using a Phenomena Identification and Ranking Tables (PIRT) informed R&D process.
\nThe first two steps of the proposed methodology are 1) Scenario Identification, where operational and accident scenarios that require analysis are indentified, and 2) PIRT, where important phenomena are identified and ranked for each scenario. In the following stages of the methodology, analysis tools are evaluated to determine whether important phenomena can be calculated and operational and accident scenarios that require study are analyzed.
\nPIRT was developed in the late 1980s, for the qualification of deterministic safety codes for Light Water Reactors [32, 33]. It is a systematic way of gathering information from experts on a specific subject, and ranking the importance of the information, in order to meet some decision-making goal [34]. An important part of the process is to also identify the associated uncertainties, usually by scoring the knowledge bases for the phenomena. Example of successful PIRT applications in thermal-hydraulics, severe accidents, fuels, materials degradation, and nuclear analysis may be found in [34]. An extensive application of PIRT is reported in [33, 35, 36], for the determination of code applicability for the analysis of selected scenarios with uncertainty evaluations.
\nIn 2007, the United States Nuclear Regulatory Commission (NRC) developed a risk-informed and performance-based regulatory structure for the licensing of future NPPs, with broader use of design specific risk information and applicable to any reactor technology [29]. Defense in Depth remains basic to this framework, providing safety margins to compensate for the uncertainties in the requirements for design, construction and operation.
\nIn this model, PSA and the Licensing Basis Events (LBEs) deterministic calculations are closely linked. LBEs are selected developing a Frequency-Consequence (F-C) curve, which is used together with the plant specific PSA.
\nThe performance-based approach is applied whenever possible, so that performance history is used to focus attention on the most important safety issues. Objective criteria are established for evaluating performance.
\nIAEA has also proposed a new safety approach and a methodology to generate technology-neutral safety requirements for advanced and innovative reactors [37]. The document identified several areas requiring further development, such as the replacement of qualitative safety objectives for quantitative ones, the enhancement of Defense in Depth, further development of PSA, the application of methodologies early in the design process, the use of an iterative design process to demonstrate the adequacy of Defense in Depth and a comprehensive review of the existing safety approach.
\nThe Objective Provisions Tree (OPT) methodology [38, 39] was suggested as a tool to systematically examine all possible options for provisions to prevent and/or control challenging mechanisms jeopardizing Defense in Depth. OPT is a systematic critical review of the Defense in Depth implementation. It identifies the required provisions that jointly ensure the prevention or control of a mechanism that represents a challenge for a safety function, which is part of a Defense in Depth Level. The existence of several challenges for each safety function and several mechanisms contributing to a given challenge leads to a tree structure. The set of provisions, jointly ensuring the prevention of one mechanism, constitutes a Line of Protection (LOP).
\nThe proposed main pillars of the new IAEA safety approach are Quantitative Safety Goals (correlated with each level of Defense in Depth), Fundamental Safety Functions, and Defense in Depth (generalized, including probabilistic considerations). Quantitative Safety Goals are based on the condition that plant states with significant frequency of occurrence have only minor or no potential radiological consequences, according to the F-C curve, called Farmer´s Curve.
\nDefense in Depth essential characteristics were described as exhaustive, balanced and graduated. “Balanced” means that no family of initiating events should dominate the global frequency of plant damage states. “Graduated” means that a progressive defense excludes the possibility of a particular provision failure to generate a major increase in potential consequences, without any possibility of recovering the situation at an intermediate stage.
\nA Risk and Safety Working Group (RSWG) was created in the frames of the Generation IV Forum to develop a safety approach for Generation IV Nuclear Systems. After its initial meeting in 2005, the first major work product, establishing the basis for the required safety approach, was released by the end of 2008 [40]. The document recommended the early application of a cohesive safety philosophy based on a concept of safety that is “built-in, not added-on”. The identification of risks must be as exhaustive as possible. The remaining lack of exhaustiveness of the accident scenarios should be covered by the notion of enveloped situations and the implementation of the Defense in Depth principles. A re-examination of the safety approach was proposed, complementing the IAEA suggestions [37] with the following attributes:
\nRisk-informed, combining both probabilistic and deterministic information.
Understandable, traceable, and reproducible.
Defensible. Whenever possible, known technology should be used.
Flexible. New information and research results should be easily incorporated.
Performance-based.
RSWG recommended a safety approach that manages simultaneously deterministic practices and probabilistic objectives, handling internal and external hazards. Besides the Defense in Depth characteristics previously mentioned (exhaustive, balanced and graduated) [37], RSWG added the followings: tolerant and forgiving. “Tolerant” means that no small deviation of a physical parameter outside the expected range can lead to severe consequences (absence of cliff-edge effects). “Forgiving” means the availability of a sufficient grace period and the possibility of repair during accident conditions.
\nRSWG recognized safety related technology gaps for the six reactor concepts selected by the Generation IV Forum in different technical areas such as updated safety approach, fuel, neutronics, thermal aerolics/hydraulics, materials & chemistry, fuel chemistry, passive safety and severe accident behavior. An effective mix of modeling, simulation, prototyping and demonstrations should be used to reduce the existing uncertainties and lack of knowledge.
\nDuring 2008 the RSWG began the development of the methodology for Generation IV safety assessments, stated in [26]. The methodology would integrate PSA and several other techniques, such as PIRT and OPT, with an extensive deterministic and phenomenological modeling [41]. In 2009 the RSWG focused its work on the development of a methodology that was denominated Integrated Safety Assessment Methodology (ISAM) [42].
\nISAM consists of five distinct analytical tools which are structured around PSA:
\nQualitative Safety Requirements/Characteristic Review (QSR).
Phenomena Identification and Ranking Table (PIRT),
Objective Provision Tree (OPT).
Deterministic and Phenomenological Analyses (DPA).
Probabilistic Safety Analysis (PSA).
From its original conception, it was clearly established that ISAM is not intended to dictate requirements or compliance with safety goals to designers. Its intention is solely to provide useful insights into the nature of safety and risk of Generation IV systems for the attainment of Generation IV safety objectives. ISAM will allow evaluation of a particular Generation IV concept or design relative to various potentially applicable safety metrics or “figures of merit” (FOM) [42, 43].
\nDuring 2010 the RSWG focused its work on the finalization of ISAM methodology presented in 2009. ISAM was conceived as a methodology providing specific tools to examine relevant safety issues at different points in the design evolution, in an iterative fashion through the development cycle. It is considered well integrated, and when used as a whole, offers the flexibility to allow a graded approach to the analysis of technical issues of varying complexity and importance [5].
\nFinally, the document describing the Generation IV Integrated Safety Assessment Methodology (ISAM) was available in June, 2011 [5]. According to this report, a principal focus of RSWG is the development and demonstration of an integrated methodology that can be used to evaluate and document the safety of Generation IV nuclear systems.
\nAn important remark is that the safety approach for Generation IV nuclear systems should differ from the one, usually applied to previous reactor generations, in which safety is generally “added on” by applying safety assessments to relatively mature designs and introducing the results in many cases as “backfits”. ISAM is therefore intended to support achievement of safety that is “built-in” rather than “added on” by influencing the direction of the concept and design development from its earliest stages [5].
\nIt is envisioned that ISAM will be a “tool kit” used in three different ways:
\nAs an integrated methodology, throughout the concept development and design phases, revealing insights capable of influencing the design evolution. ISAM can develop a more detailed understanding of safety-related design vulnerabilities, and resulting risk contributors. New safety provisions or design improvements can be identified, developed, and implemented relatively early.
Applying selected elements of the methodology separately at various points throughout the design evolution to yield an objective understanding of safety-related issues (such as risk contributors, safety margins, effectiveness of safety provisions, uncertainties, etc.) important to decision makers.
In the late stages of design maturity, for decision makers and regulators to measure the level of safety and risk associated with a given design relative to safety objectives or licensing criteria (“post facto” application).
ISAM is essentially a PSA-based safety assessment methodology, with the additional strength of other tools, tailored to answering specific types of questions at various stages of design development. The methodology is well integrated. Although individual analytical tools can be selected for separate and exclusive use, the full value of the integrated methodology is derived from using each tool, in an iterative fashion and in combination with the others, throughout the development cycle. Figure 7 shows the overall task flow of ISAM and indicates which tools are intended for each phase of Generation IV system technology development [5].
\nISAM Task Flow [5].
The methodology comprises several stages, from the pre-conceptual design to licensing and operation, with the corresponding safety features and criteria moving from primarily qualitative to quantitative. In the early design stages, the main role corresponds to the qualitative safety analysis techniques QSR, PIRT and OPT, but Deterministic and Phenomenological Analysis (DPA) and Probabilistic Safety Assessment (PSA) are introduced earlier in comparison with the current practice for previous reactor generations. DPA and PSA are the key techniques to be applied during the last stages corresponding to final design, licensing and operation, where safety criteria are mainly quantitative.
\nOnly one of the tools integrated in ISAM is completely new, and was developed specially for Generation IV Reactor Systems: the Qualitative Safety features Review (QSR). QSR is intended to provide a systematic means of ensuring and documenting that the evolving Generation IV system concept of design incorporates the desirable safety-related attributes and characteristics that are identified and discussed in [40].
\nQSR is conducted using a template structure organized according to the first four levels of Defense in Depth (prevention, control, protection and management of severe accidents). The review is based on an exhaustive check list of safety good practices and recommendations applicable to Generation IV systems. Design options are evaluated, identifying their strength or weakness, and qualified as favorable, unfavorable or neutral in relation with the desirable characteristics.
\nAn exhaustive and detailed check list is essential for the QSR quality. To generate the list, a top-down functional approach is conducted, as shown in Figure 8 [5]. Firstly, the recommendations from RSWG, IAEA standards, INSAG and INPRO guidelines and other references, organized according to the levels of Defense in Depth, are detailed as much as possible, for a technology neutral condition. Finally, the defined neutral characteristics are used to develop the set of specific recommendations for a particular technology. In this process, the characteristics and features are grouped in four classes, moving from general recommendations to detailed specific attributes [5]:
\nClass 1: Generic and Technology neutral.
\nClass 2: Detailed and Technology neutral.
\nClass 3: Detailed and Technology neutral but applicable to a given safety function.
\nClass 4: Detailed and applicable to a given safety function and specific technology.
\nRSWG has developed check lists covering desirable Classes 1 to 3 safety characteristics for the first four levels of Defense in Depth. Class 3 recommendations were determined for the safety function “Decay Heat Removal” [5]. It is considered that QSR applications will be important for the identification of Generation IV R&D needs.
\nTop-down approach to establish safety recommendations derived from a QSR [5]
In ISAM, PIRT and OPT are tools that complement each other. PIRT is used to identify phenomena impacting accident scenarios while OPT finds out the necessary Defense in Depth provisions to prevent, control or mitigate the corresponding phenomena. They can be iteratively applied from conceptual to mature designs. Both techniques are basically qualitative tools based on expert elicitation.
\nPIRT is more expert-dependant due to the complexity and diversity of possible phenomena involved in accident scenarios. The expert panel selects a FOM, for example, the fission products release, the maximum core coolant temperature, etc.; and the PIRT process is applied to each of the identified accident scenarios to determine and categorize the safety-related phenomena and the uncertainties in their knowledge. Usually a four level scale is used to rank the phenomena importance (high, medium, low or insignificant) and existing knowledge (full, satisfactory, partial or very limited). The ranked values of importance and uncertainty are useful to determine R&D effort priorities for accident scenarios. Safety issues located in the region of high importance and large uncertainty have the greatest priorities.
\nOPT is a systematic and structured top-down method for a fully characterization of the Defense in Depth architecture by identifying in detail the set of objective realistic countermeasures against a great diversity of safety-deteriorating mechanisms. Following a deductive approach, the model goes from each level of defense throughout the safety objectives and physical barriers; down to the safety functions and their challenges, until challenging mechanisms at the lowest tree levels are deduced and the associated sets of objective provisions in opposition to them, constituting LOPs, are determined. Figure 9 shows the tree structure of the OPT process [39].
\nOPT process to determine Defense in Depth objective safety provisions [39]
The application of OPT is expected to contribute from early stages to the built-in safety approach proclaimed as part of Generation IV safety assessment philosophy. Its final objective is to ensure that Defense in Depth satisfies the desirable attributes of being exhaustive, balanced, progressive, tolerant and forgiving [5]. During the pre-conceptual and conceptual design phases OPT will serve as an important guidance to research efforts in order to achieve the mentioned Defense in Depth attributes by means of robust, reliable and simple design solutions. OPT can also identify degradation mechanisms representing feedbacks to PIRT.
\nPSA is the principal basis of ISAM, but DPA also constitutes a vital component, providing support to PSA, as well as to PIRT and OPT. Following the Risk-informed methodology, ISAM integrates PSA, DPA and Defense in Depth (evaluated using the OPT). The two mutually complemented methodologies, PSA and DPA, contribute to assess important phenomena identified in the PIRT process, and can also evaluate the effectiveness of LOPs deduced by means of OPT. DPA is indispensable to understand NPP safety issues, providing quantitative insights, important as PSA inputs. Best-estimated deterministic computer codes are preferred, incorporating sensitivity analyses to cover the existing uncertainties, depending on the design stage. One important challenge is the upgrade, development and validation of deterministic computer codes and their necessary input data to perform convincing safety assessments of some innovative reactor concepts.
\nRSWG supports the idea of applying PSA from the earliest practical stages of the design process, and continuing its application iteratively throughout the evolution of the design concept until its maturity, in the stages of final design, licensing and operation. The PSA scope should comprise both internal and external events. PSA is recognized as a fundamental tool to prioritize properly design and operational issues which are more significant to safety, contributing in this way to a proper balance between costs and safety effectiveness of Generation IV Nuclear Energy Systems. PSA advantages for a systematic understanding and evaluation of risk uncertainties is also remarked. It is expected that PSA will contribute to understand differences in the level of safety of diverse technical proposals and select the designs that better fulfill the selection safety criteria for a given Generation IV reactor concept.
\nThe traditional PSA metrics of Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) remain useful, although the first one does not apply to all the new reactor concepts. It has been recommended the generalization of CDF as “undesirable event with significant source term mobilization”. The principal risk metric that should be used for comparing Generation IV concepts is the Farmer’s Curve, oriented to the Generation IV Safety objective regarding the elimination of off-site emergency response [5]. To achieve its leading role, PSA must be validated as part of a rigorous quality assurance program, including a peer review conducted by independent experts.
\nISAM is a relatively new methodology which is still under development and adjustment. It is recognized that the methodology will need to be modified or updated based on the lessons and findings derived from the Fukushima accident [5, 44].
\nAn example of a preliminary ISAM application to the Japanese Sodium-cooled Fast Reactor (JSFR) is described in [5]. A PIRT process was conducted for the reactor shutdown by the passive Self-Actuated Shutdown System (SASS) upon an Unprotected Loss of Flow (ULOF) accident. OPT results are presented for the safety function “Core heat removal” of Defense in Depth Level 3, indentifying several mechanisms for the challenge “Degraded or disruption of heat transfer path” and their corresponding LOPs. Applicability of DPA and PSA is also shown.
\nRSWG demonstration that ISAM can be used to evaluate and document the safety of Generation IV nuclear systems, supported by an extended use of the methodology in practical applications, is only beginning but has excellent perspectives of becoming a future reality. It will certainly depend on international efforts and progress in the materialization of Generation IV reactor concepts. RSWG has been asked for GIF to work toward the provision of increasingly detailed guidance for application of ISAM in the development of Generation IV systems [44].
\nThe economic modeling working group (EMWG) was formed in 2004 for developing a cost estimating methodology to be used for assessing Generation IV International Forum (GIF) systems against its economic goals. Its creation followed the recommendations from the economics crosscut group of the Generation IV roadmap project that a standardized cost estimating protocol be developed to provide decision makers with a credible basis to assess, compare, and eventually select future nuclear energy systems, taking into account a robust evaluation of their economic viability. The methodology developed by the EMWG is based upon the economic goals of Generation IV nuclear energy systems, as adopted by GIF: to have a life cycle cost advantage over other energy sources, to have a level of financial risk comparable to other energy projects (i.e., to involve similar total capital investment and capital at risk).
\nThis section briefly describes an economic model for Generation IV nuclear energy systems [6] and the accompanying software [45] in which the guidelines and models were implemented. These tools will integrate cost information prepared by Generation IV system development teams during the development and demonstration of their concept, thus assuring a standard format and comparability among concepts. This methodology will allow the Generation IV International Forum (GIF) Experts Group to give an overview to policy makers and system development teams on the status of available economic estimates for each system and the relative status of the different systems with respect to Generation IV economic goals. Figure 10 displays the structure of the integrated nuclear energy economic model. The following discussion is based on this figure.
\nStructure of the integrated nuclear energy economic model [6].
The model is split in four parts: construction/production, fuel cycle, energy products, and modularization.
\nCost estimates prepared by system design teams should report the overall direct and indirect costs for reactor system design and construction (base construction cost) and an estimate of the reactor annual operation and maintenance costs. The intent is that these costs be developed using the GIF COA described in [6], prepared by the methods outlined therein. The decision maker, however, needs more than just the overall costs in each life cycle category. Of particular interest are the cost per kilowatt of installed capacity and the cost of electricity generation (cost per kilowatt-hour) from such systems, including the contribution of capital and non-fuel operations.
\nRef. [6] describes how interest during construction (IDC), contingencies, and other supplemental items are added to the base construction cost to obtain the total project capital cost. This total cost is amortized over the plant economic life so that the capital contribution to the levelized unit of energy cost (LUEC) can be calculated. Operation and maintenance (O&M) and decontamination and decommissioning (D&D) costs, along with electricity production information, yield the contributions of non-fuel costs to the overall cost of electricity. These algorithms have been derived from earlier Oak Ridge National Laboratory (ORNL) nuclear energy plant databases (NECDB) [46, 47], to calculate these costs.
\nFuel cycle materials and services are purchased separately by the utility or the fuel subcontractor. For fuel cycles commercially deployed, there are mature industries worldwide that can provide these materials and services. Markets are competitive, and prices are driven by supply and demand. The fuel cycle model requires as inputs the amount of fuel needed for the initial core and subsequent equilibrium cores, along with the fissile enrichment of the uranium or plutonium, and, for uranium, the transaction tails assay assumed by the enrichment service provider. The EMWG model uses algorithms similar to those described in [48] to estimate the overall cost for each step and ultimately the unit cost contribution of fuel to electricity cost. Background material on the economic aspects of fuel cycle choices including information on nuclear materials and fuel cycle service unit costs for conventional reactor types that use commercially available fuels can be found in NEA reports [48, 49]. These documents include cost data on fuel reprocessing and high-level waste disposal for closed fuel cycles and spent fuel disposal for the once-through option.
\nInnovative fuel cycles or fuel cycle steps for which no industrial scale or commercial facilities currently exist, especially for fuel fabrication, reprocessing, and waste disposal are also addressed. For example, the Very-High-Temperature Reactor system will require high-temperature particle fuel and the SFR (Sodium-Cooled Fast Reactor) system might require innovative pyrometallurgical and pyrochemical facilities for fuel fabrication, reprocessing, and re-fabrication. For such systems, price data for fuel cycle services generally are not readily available. Therefore, a unit cost of fuel cycle services, such as $/kgHM (heavy metal) for fuel fabrication, should be calculated using a methodology similar to that used for LUEC calculation for the reactor system. The design team must supply data on the design and construction costs for the facilities, along with an estimate of their annual production rates and operation costs. Algorithms discussed in [6] can produce rough approximations of the unit costs.
\nThe heat generated by some Generation IV systems has the potential for uses other than electricity generation, such as the production of hydrogen by thermal cracking of steam. There are also possible co-production models where the heat is used for both electricity production and process heat applications. The energy products model deals with these issues and is also discussed.
\nCost issues and possible economic benefits that might result from modularization or factory production of all or part of a reactor system are also discussed [6].
\nThe aim is to furnish a standardized cost estimating protocol to provide decision makers with a credible basis to assess, compare, and eventually select future nuclear energy systems taking into account a robust evaluation of their economic viability. To provide a credible, consistent basis for the estimated costs, early estimates of the evolving design concepts are expected to be based on conventional construction experience of built plants. This limitation is desirable from a consistency point of view because it can provide a reasonable starting point for consistent economic evaluation of different reactor concepts.
\nOf particular interest in this sense is the work by Hejzlar et al [50], related to the Gas Cooled Fast Reactor (GFR). They discuss challenges posed by the GFR when striving for the achievement of balance among the Generation IV goals. According to Carelli et al [51], the nuclear option has to face not only the public opinion sensibility, mainly related to plant safety and waste disposal issues, but also the economic evaluation from investors and utilities, particularly careful on that energy source and in deregulated markets. Smaller size nuclear reactors can represent a viable solution especially for developing countries, or countries with not-highly-infrastructured and interconnected grids, or even for developed countries when limitation on capital at risk applies. A description of Small-Medium size Reactor (SMR) economic features is presented, in a comparison with the state-of-the-art Large size Reactors. A preliminary evaluation of the capital and operation and maintenance (O&M) costs shows that the negative effects of the economies of scale can be balanced by the integral and modular design strategy of SMRs.
\nTechnical and institutional characteristics of Generation IV systems are used to evaluate system response and determine its resistance against proliferation threats and robustness against sabotage and terrorism. System response outcomes are expressed in terms of a set of measures.
\nThe methodology is organized to permit evaluations to be performed at the earliest stages of system design and to become more detailed and more representative as design evolves. Uncertainty of results is incorporated into the evaluation.
\nThe results are intended for three types of users: system designers, program policy makers, and external stakeholders. Program policy makers will be more likely to be interested in the high-level results that discriminate among choices, while system designers and safeguards experts will be more interested in results that directly relate to design options that will improve their performance (e.g., safeguards by design).
\nThe proliferation resistance and physical protection Working Group has based its specification of the evaluation methodology scope on the definition of the Generation IV proliferation resistance and physical protection goal. The Generation IV Technology Roadmap (DOE, 2002b) [52] formally defined the following proliferation resistance and physical protection goal for future nuclear energy systems:
\n\n Generation IV nuclear energy systems will increase the assurance that they are a very unattractive and the least desirable route for diversion or theft of weapons-usable materials, and provide increased physical protection against acts of terrorism.\n
\nThe definition of proliferation resistance adopted by the Working Group agrees with the definition established at the international workshop sponsored by the International Atomic Energy Agency (IAEA, 2002b)[53].
\nFormal definitions of proliferation resistance and physical protection have been established as presented next.
\nProliferation resistance is that characteristic of a nuclear energy system that prevents the diversion or undeclared production of nuclear material and the misuse of technology by the host state seeking to acquire nuclear weapons or other nuclear explosive devices.
\nPhysical protection (robustness) is that characteristic of a nuclear energy system that prevents the theft of materials suitable for nuclear explosives or radiation dispersal devices and the sabotage of facilities and transportation by sub-national entities or other non-host state adversaries.
\nThe proliferation resistance and physical protection technology goal for Generation IV nuclear energy systems, when combined with the definitions of proliferation resistance and physical protection, is therefore as follows.
\nA Generation IV nuclear energy system is to be the least desirable route to proliferation by hindering the diversion of nuclear material from the system and hindering the misuse of the nuclear energy system and its technology in the production of nuclear weapons or other nuclear explosive devices.
\nA Generation IV nuclear energy system is to provide enhanced protection against theft of materials suitable for nuclear explosives or radiation dispersal devices and enhanced protection against sabotage of facilities and transportation. The proliferation resistance and physical protection methodology provides the means to evaluate Generation IV nuclear energy systems with respect to the following categories of proliferation resistance and physical protection threats:
\nProliferation Resistance – Resistance to a host state’s acquisition of nuclear weapons by:
\nConcealed diversion of material from declared flows and inventories;
Overt diversion of material from declared flows and inventories;
Concealed material production or processing in declared facilities;
Overt material production or processing in declared facilities;
Concealed material production or processing by replication of declared equipment in clandestine facilities.
Physical Protection (robustness)
Theft of nuclear weapons-usable material or information from facilities or transportation;
Theft of hazardous radioactive material from facilities or transportation for use in a dispersion weapon (radiation dispersal device or “dirty bomb”);
Sabotage at a nuclear facility or during transportation with the objective to release radioactive material to harm the public, damage facilities, or disrupt operations.
\n Figure 11 illustrates the most basic methodological approach. For a given system, analysts define a set of challenges, analyze system response to these challenges, and assess outcomes. The challenges to the nuclear energy system are the threats posed by potential proliferate states and by sub-national adversaries. The technical and institutional characteristics of Generation IV systems are used to evaluate the system response and determine its resistance to proliferation threats and robustness against sabotage and terrorism. The outcomes of the system response are expressed in terms of proliferation resistance and physical protection measures.
\nBasic framework for the proliferation resistance and physical protection evaluation methodology [7]
The evaluation methodology assumes that a nuclear energy system has been at least conceptualized or designed, including both intrinsic and extrinsic protective features of the system. Intrinsic features include the physical and engineering aspects of the system; extrinsic features include institutional aspects such as safeguards and external barriers. A major thrust of the proliferation resistance and physical protection evaluation is to elucidate the interactions between intrinsic and extrinsic features, study their interplay, and then guide the path toward an optimized design.
\nThe structure for the proliferation resistance and physical protection evaluation can be applied to the entire fuel cycle or to portions of a nuclear energy system. The methodology is organized as a progressive approach to allow evaluations to become more detailed and more representative as system design evolves. Proliferation resistance and physical protection evaluations should be performed at the earliest stages of design when flow diagrams are first developed in order to systematically integrate proliferation resistance and physical protection robustness into the designs of Generation IV nuclear energy systems along with the other high-level technology goals of sustainability, safety and reliability, and economics. This approach provides early, useful feedback to designers, program policy makers, and external stakeholders from basic process selection (e.g., recycling process and type of fuel), to detailed layout of equipment and structures, to facility demonstration testing.
\n\n Figure 12 provides an expanded outline of the methodological approach. The first step is threat definition. For both proliferation resistance and physical protection, the threat definition describes the challenges that the system may face and includes characteristics of both the actor and the actor’s strategy. For proliferation resistance, the actor is the host state for the nuclear energy system, and the threat definition includes both the proliferation objectives and the capabilities and strategy of the host state. For physical protection threats, the actor is a sub-national group or other non-host state adversary. The physical protection actors’ characteristics are defined by their objective, which may be either theft or sabotage, and their capabilities and strategies.
\nFramework for the proliferation resistance and physical protection evaluation methodology [7]
The proliferation resistance and physical protection methodology does not determine the probability that a given threat might or might not occur. Such evaluations may come from national threat evaluation organizations. The proliferation resistance and physical protection evaluation is based on design features of facilities as well as institutional considerations. Therefore, the selection of what potential threats to include is performed at the beginning of a proliferation resistance and physical protection evaluation, preferably with input from a peer review group organized in coordination with the evaluation sponsors. The uncertainty in the system response to a given threat is then evaluated independently of the probability that the system would ever actually be challenged by the threat. In other words, proliferation resistance and physical protection evaluations are challenge dependent.
\nThe detail with which threats can and should be defined depends on the level of detail of information available about the nuclear energy system design. In the earliest stages of conceptual design, where detailed information is likely limited, relatively stylized but reasonable threats must be selected. Conversely, when design has progressed to the point of actual construction, detailed and specific characterization of potential threats becomes possible.
\nWhen threats have been sufficiently detailed for the particular evaluation, analysts assess system response, which has four components:
\nSystem Element Identification. The nuclear energy system is decomposed into smaller elements or subsystems at a level amenable to further analysis. The elements can comprise a facility (in the systems engineering sense), part of a facility, a collection of facilities, or a transportation system within the identified nuclear energy system where acquisition (diversion) or processing (proliferation resistance) or theft/sabotage (physical protection) could take place.
Target Identification and Categorization. Target identification is conducted by systematically examining the nuclear energy system for the role that materials, equipment, and processes in each element could play in each of the strategies identified in the threat definition. Proliferation resistance targets are nuclear material, equipment, and processes to be protected from threats of diversion and misuse. Physical protection targets are nuclear material, equipment, or information to be protected from threats of theft and sabotage. Targets are categorized to create representative or bounding sets for further analysis.
Pathway Identification and Refinement. Pathways are potential sequences of events and actions followed by the actor to achieve objectives. For each target, individual pathways are divided into segments through a systematic process, and analyzed at a high level. Segments are then connected into full pathways and analyzed in detail. Selection of appropriate pathways will depend on the scenarios themselves, the state of design information, the quality and applicability of available information, and the analyst\'s preferences.
Estimation of Measures. The results of the system response are expressed in terms of proliferation resistance and physical protection measures. Measures are the high-level characteristics of a pathway that affect the likely decisions and actions of an actor and therefore are used to evaluate the actor’s likely behavior and outcomes. For each measure, the results for each pathway segment are aggregated as appropriate to compare pathways and assess the system so that significant pathways can be identified and highlighted for further assessment and decision making
The measures for proliferation resistance are:
\nProliferation Technical Difficulty – The inherent difficulty, arising from the need for technical sophistication and materials handling capabilities, required to overcome the multiple barriers to proliferation.
Proliferation Cost – The economic and staffing investment required to overcome the multiple technical barriers to proliferation, including the use of existing or new facilities.
Proliferation Time – The minimum time required to overcome the multiple barriers to proliferation (i.e., the total time planned by the Host State for the project).
Fissile Material Type – A categorization of material based on the degree to which its characteristics affect its utility for use in nuclear explosives.
Detection Probability – The cumulative probability of detecting a proliferation segment or pathway.
Detection Resource Efficiency – The efficiency in the use of staffing, equipment, and funding to apply international safeguards to a nuclear energy system.
The measures for physical protection are:
Probability of Adversary Success – The probability that an adversary will successfully complete the actions described by a pathway and generate a consequence.
Consequences – The effects resulting from the successful completion of the adversary’s action described by a pathway.
Physical Protection Resources – The staffing, capabilities, and costs required to provide PP, such as background screening, detection, interruption, and neutralization, and the sensitivity of these resources to changes in the threat sophistication and capability.
By considering these measures, system designers can identify design options that will improve system proliferation resistance and physical protection performance. For example, designers can reduce or eliminate active safety equipment that requires frequent operator intervention.
\nThe final steps in proliferation resistance and physical protection evaluations are to integrate the findings of the analysis and to interpret the results. Evaluation results should include best estimates for numerical and linguistic descriptors that characterize the results, distributions reflecting the uncertainty associated with those estimates, and appropriate displays to communicate uncertainties.
\nFurther literature on the subject of this section comprises, for example, the paper by Bari et al [54], where the general methodology for proliferation resistance and physical protection is discussed and applied. An application to an example sodium fast reactor is discussed in terms of elicitation in [55]. An application concerning nuclear fuel cycles is discussed in [56]. A practical tool to assess proliferation resistance of nuclear energy systems is discussed in [57].
\nProliferation resistance is discussed in [58] concerning the mobile fuel reactor, which is not one of Generation IV concepts in discussion, but interesting insights may be found therein. Penner et al [59] discuss new reactor designs and construction where a Generation IV design perspective is presented and proliferation resistance is set as an issue of utmost importance. Lennox et al [60] discuss the plutonium issue from the point of view of Generation IV designs.
\nThe molten salt reactor is focused on proliferation issues in Ref. [61] also. Here proliferation considerations are discussed in face of the reactor operation because without the removal of plutonium and uranium from the fuel mixture, the reactivity starts to fluctuate and needs compensation. Uri and Engel [62] discuss non-proliferation attributes of molten salt reactors, as, for example, less plutonium stocks.
\nMyths of the proliferation resistance approach are focused in Ref. [63].
\nThe discussion presented in this chapter clearly shows that much effort has been developed on a worldwide basis for conceiving the reactors that will be in use around 2030. Due to its beginning as a military weapon, nuclear energy is not an energy option that is accepted without strong restrictions in many countries. This resistance has been particularly aggravated immediately after the accidents in Three Mile Island, Chernobyl and the recent one in Fukushima. Many lessons learned from these accidents have been employed in the conception of Generation IV reactors, as many of them had already been implemented in Generation III reactors, like Westinghouse\'s AP1000.
\nThe Generation IV philosophy for reactor development brings into light concerns about sustainability, economic viability, safety, and security translated into the concepts of proliferation resistance and physical protection. These are new concepts that are playing the dominante roles in reactor development for the future. It is also noteworthy that safety analysis is to be stressed, mainly the application of the risk-informed decision making approach for licensing purposes. Certainly, this integrated phylosophy will do much for turning nuclear energy systems much more acceptable by the final users.
\nIn a rapidly growing world population and toward meeting consumers’ needs, solid waste landfills will continue receiving huge volumes of waste. Therefore, waste management is becoming increasingly mandatory for the promotion of environmental sustainability. Numerous regulations have been imposed worldwide by governments and environmental organizations in order to reduce the negative environmental impact resulting from large numbers of solid waste landfills. The transformation of a large amount of solid waste into an alternative resource will preserve the reducing nonrenewable resources of materials; maintain the required energy and also will help solve environmental and exhausted landfill problems. Until today, researchers are investigating new solid waste materials and the potentials of recycling either in other industries or new products.
Being the world’s most consumed human-made material, concrete attracted considerable interest as a possible way to recycle solid waste products especially those that can replace cement which is a significant contributor to global greenhouse gas emissions. An equal amount of CO2 is generated for the production of Portland cement [1]. The cement industry produces around 5–8% of the annual global greenhouse gas emissions released into the atmosphere [2]. Several by-products such as fly ash, slag, and silica fume are effectively being used in the daily production of concrete as partial cement replacement (i.e., supplementary cementitious materials (SCM)) to reduce CO2 emission [3, 4].
Global production of ceramic tiles is more than 12 Billion m2 [5]. The manufacture of ceramic tiles generates ceramic waste powder (CWP) during the final polishing process at a rate of 19 kg/m2 [6]. Therefore, the global generation of CWP exceeds 22 Billion tons. The CWP represents a significant challenge to get rid of concerning its environmental impact. It can cause, soil, water, and air pollution. On the other hand, it could represent an excellent opportunity to be used as an alternative concrete ingredient if it could be utilized in making concrete.
The effect of using ceramic wastes (i.e., roof tiles, blocks, bricks, electrical insulators, etc.) as aggregates or SCM in conventional-vibrated concrete (CVC) and mortar was reported in several studies. It is noted that limited studies were conducted on using CWP as a cement replacement in self-compacting concrete (SCC) and alkali-activated concrete (AAC) (i.e., geopolymer concrete). Some studies investigated the use of ceramic waste as coarse aggregates in CVC and mortar [7, 8, 9, 10, 11, 12, 13, 14, 15, 16]. It was concluded that ceramic waste could be used as partial replacement of natural coarse aggregate. The ceramic waste aggregate should be pre-saturated by water to offset its high absorption. The compressive strength decreased if the ceramic waste replaced natural coarse aggregate beyond 25% by weight. The use of ceramic waste as fine aggregate in CVC and mortar was assessed by various researchers [16, 17, 18, 19, 20, 21, 22]. It was noted that using a high content of ceramic waste as fine aggregate had a negative impact on the workability of the fresh concrete, and workability admixtures were needed to avoid any adverse effect on concrete workability. It was concluded that the use of 50% by weight replacement of fine natural aggregate by ceramic waste could produce concrete without affecting the performance of hardened concrete.
The use of CWP as partial replacement of cement attracted the attention of several researchers [6, 23, 24, 25, 26, 27, 28, 29, 30, 31, 32, 33, 34, 35]. The main conclusion from the studies was that CWP showed slow pozzolanic activity which was evidenced at late ages. The early compressive strength was reduced by the inclusion of CWP. The development of compressive strength needed time. On the other hand, durability was improved by the incorporation of CWP in the mixtures. It was noticed that the investigations on using CWP as partial replacement of cement did not address the fresh concrete properties as affected by the inclusion of CWP as well as the microstructure characteristics. Also, no guidelines were provided for using CWP to partially replace cement. The CWP replacement level will depend on personal knowledge and experience. Furthermore, the replacement of cement by large quantities of CWP needs further evaluation.
The use of CWP in self-compacting concrete (SCC) mixtures received limited attention. In 2017, Subaşi et al. [36] investigated the use of CWP as a partial cement replacement in SCC mixtures. It was concluded that CWP could replace 15% by weight of the cement without adversely affecting the properties of the produced SCC. In 2018, Jerônimo et al. [37] replaced cement by ground clay brick waste (GCBW) in SCC mixtures. It was concluded that 20–30% by weight of the cement could be replaced by GCBW, and the compressive strength improved at 90 days of age. It was observed that the detailed evaluation of the SCC fresh properties as affected by the inclusion of CWP was not addressed. Also, the effect of using high-volume CWP in SCC still needs further assessment.
Concerning using CWP in alkali-activated concrete (AAC) (i.e., geopolymer concrete), it was noted that very limited investigations were conducted [38, 39, 40]. The main conclusion that CWP could be used in making AAC but needs detailed investigation and assessment.
An in-depth investigation to study the utilization of CWP in the production of different types of concrete is needed. This chapter summarizes the findings of collective studies conducted by the authors investigating the use of CWP in making eco-friendly concrete [41, 42, 43, 44, 45], with a particular focus on using CWP as a partial cement replacement in CVC and SCC, and the production of AAC. This will establish better understanding on how to incorporate an existing solid waste as a new construction ingredient in making echo-friendly concretes in order to optimize solid waste management, and help protect the environment by reducing the use of cement and efficiently getting rid of a solid waste material.
The produced ceramic waste material was a wet material due to the use of water during the polishing process. The average moisture content was 36% by mass. The average specific surface area (SSA) measured by air-permeability (i.e., Blain air permeability test apparatus) was 555 m2/kg. More than 50% by volume of the CWP particles had a size ranging between 5 and 10 μm. Figure 1 shows the particles’ size distribution of the CWP.
Particle size distribution of CWP [43]. Reproduced with permission from the publisher.
The CWP consisted of irregular and angular particles which are similar to cement particles in shape as shown in the scanning electron microscope (SEM) image in Figure 2. Figure 3 shows the energy dispersive spectroscopy (EDS) of the main oxides of the CWP. The EDS analysis indicated that CWP is mainly composed of SiO2 and Al2O3.
SEM images of CWP.
EDS analysis of CWP [43]. Reproduced with permission from the publisher.
Table 1 gives the chemical analysis of the CWP as determined by X-ray fluorescence (XRF). CWP is mainly composed of silica (SiO2) and alumina (Al2O3). Both oxides are around 85% of the total material mass. Other compounds (i.e., CaO, MgO, and SO3) exist in small quantities. The mass fractions of (SiO2 + Al2O3 + Fe2O3) satisfies the requirement of the ASTM C618 [46] for natural pozzolana (i.e., >70%). Also, the SO3 and the loss on ignition (L.O.I.) conformed to the ASTM C618 requirements.
CaO | SiO2 | Al2O3 | MgO | Fe2O3 | SO3 | L.O.I. |
---|---|---|---|---|---|---|
1.70(0.69) | 68.60(0.97) | 17.00(0.57) | 2.50(0.90) | 0.80(0.04) | 0.12(0.16) | 1.78 |
Chemical composition of CWP using XRF (modified from [43]).
Note: Values in parentheses are the standard deviation.
Figure 4 displays the X-ray diffraction (XRD) analysis of the CWP. The XRD indicates that the main peaks were noticed between 2-theta values of 20 and 30o which indicates the presence of (SiO2). The observed hump between 20 and 30o indicates the occurrence of an amorphous phase. Moreover, the unleveled graph trend between the 2-theta values 0 and 40o indicates the existence of an amorphous phase in the CWP sample.
XRD pattern of CWP [43]. Reproduced with permission from the publisher.
Characterizing industrial waste materials and their potentials is one of the challenging issues in the field of cement and concrete. The compressive strength was given prominence as an initial means for evaluating the pozzolanic activity. The compressive strength development of cement mortar including CWP is assessed according to ASTM C311 [47] to measure the strength activity index (SAI).
Four mortar mixtures are prepared in which cement is partially replaced by CWP. The replacement levels are 10, 20, 30 and 40% by weight. Strength activity index (SAI) is calculated as the strength percentage as compared to the control mortar mixture. Table 2 gives the 28 days compressive strength, standard deviation SAI. Results showed that all CWP specimens satisfied the ASTM C618 requirement of SAI (i.e., >75%). In an investigation by Steiner et al. [25], a similar trend in the activity index for mortar mixtures with ceramic tiles polishing residues was reported. The SAI decreased after the inclusion of 40% CWP by cement mass; this could be attributed to the dilution effect. Also, it might be due to the high silica available in the mixture as a result of the high CWP. This large quantity could not find sufficient calcium hydroxide (CH) in order to react with. Therefore, most of the silica components were left without getting involved in the chemical reaction [48]. Also, Frattini test [49] is performed to identify the pozzolanic activity of CWP following BS EN 196-5:2011 [50]. Test samples with 0, 20 and 40% CWP as cement replacement by weight are tested. The Frattini test showed that concrete with 20 and 40% CWP replacement of Portland cement exhibited pozzolanic activity at 8 and 28 days age of concrete as shown in Figure 5.
CWP replacement level (mass %) | ||||
---|---|---|---|---|
10% | 20% | 30% | 40% | |
Average 28 days strength (MPa) | 39.9 | 46.0 | 48.8 | 37.5 |
Standard deviation (MPa) | 4.0 | 3.0 | 4.4 | 1.2 |
Strength activity index (SAI) in (%) | 91.0 | 105.0 | 110.5 | 85.5 |
Strength activity index (SAI) results for CWP [43].
Reproduced with permission from the publisher.
Frattini test at 8 and 28 days of CP with CWP replacement [45]. Reproduced with permission from the publisher.
In conclusion, CWP is silica and alumina rich material with some amorphous phases. The CWP has some pozzolanic activity, especially at a late age, as confirmed by strength activity index and Frattini tests. Therefore, CWP possesses the potentials to be used as a partial cement replacement in CVC and SCC mixtures, and as a main binder source to make AAC mixtures.
CWP is used to partially replace cement (0, 10, 20, 30 and 40% by weight) in different CVC mixtures. Two concrete grades with different cement contents are studied (25 and 50 MPa). The mixtures are chosen to cover several applications and different cement contents. All mixtures are designed to have a slump value from 60 to 100 mm. Table 3 gives the mixtures’ proportions of the mixtures. Initial slump values (i.e., ASTM C 143 [51]) is used to judge the mixtures’ workability. The time to reach zero slump is used to assess the workability retention of the concrete mixtures. The development of compressive strength with age (i.e., 7, 28 and 90 days) and drying shrinkage (i.e., 120 days) are measured. Rapid chloride ion penetration test (RCPT) (i.e., ASTM C 1202 [52]) and bulk electrical resistivity test (i.e., ASTM C 1760 [53]) are conducted at 28 and 90 days of age to evaluate the durability of the concrete mixtures. Triplicate samples are used for the compressive strength, drying shrinkage, RCPT, bulk electrical resistivity and permeable pores tests and the average results are used. The development of the microstructure is assessed by measuring permeable pores (i.e., ASTM C642 [54]) and the pore system (i.e., total porosity and median pore diameter) is measured by mercury intrusion porosimetry (MIP). Both are measured at 90 days of age. Main microstructure characteristics are identified using scanning electron microscopy (SEM).
Concrete mixtures are prepared using ordinary Portland cement (OPC) as the primary binder. The specific surface area of cement is 380 m2/kg. Natural crushed stone of maximum size 19.0 mm is used as coarse aggregate. The specific gravity is 2.65 while the absorption was 1%. Natural sand with fineness modulus between 2.5 and 2.7 is used as fine aggregate. The specific gravity is 2.63.
Initial slump values are given in Table 3. As CWP inclusion level increases, the initial slump value decreases as a result of its high specific surface area (SSA) compared to that of the cement (i.e., the SSA of CWP is 1.5 times that of the cement). Workability retention defines the time available for easy handling the mixture. Figure 6 shows the time to zero slump of the concrete mixtures including CWP. It is noted that the workability retention time increases due to the inclusion of CWP. This could a result of CWP has no hydraulic reaction, and its pozzolanic reaction is slow. The use of 10% CWP in the 25 MPa mixtures has the highest workability retention. While for the 50 MPa mixtures, the use of 20% CWP shows the best retention time.
Time to zero slump.
The compressive strength development at different ages is shown in Figure 7. The coefficient of variation (COV) ranged from 0.4 to 4.8%. The compressive strength values at 7 and 28 days of age are lower than the target strength for both mixtures (i.e., 25 and 50 MPa). The reduction in strength is proportional to the CWP content. This could be attributed to the fact that CWP has no hydraulic reaction. Also, its contribution to early strength depended mainly on its microfilling ability (i.e., CWP particles’ size ranged from 5 to 10 μm). This behavior agrees with that of most pozzolanic materials with slow strength development at early ages [55]. Also, slowed strength development at early ages is reported for CWP [28, 29, 30, 32].
Compressive strength development with age.
At a late age (i.e., 90 days) all the 25 MPa mixtures including CWP achieve compressive strength values higher than the target strength. The mixture with 10% CWP shows the highest compressive strength. The strength gain at 90 days of age might be due to the pozzolanic characteristics of the CWP material. For the 50 MPa mixtures, all CWP mixtures the target strength is achieved. The increase in strength values could be justified by the delayed pozzolanic reaction of the CWP. The CWP particles could have worked as nucleation sites for cement grains and hydration products which led to a denser microstructure.
Table 4 shows the 120 days drying shrinkage strain values. The COV ranged from 20 to 26%. It is observed that the drying shrinkage strain decreases with increasing the CWP replacement level. The pores’ structure and connectivity of pores are changed due to the fine CWP particles and its pozzolanic action. This change results in restricting water movement through the concrete. The drying shrinkage values for mixtures including 10 and 20% CWP do not differ significantly from that of the control mixtures. For the 25 MPa mixtures, CWP with replacement levels of more than 20% reduces the drying shrinkage strain between 29 and 60% compared to the control mixture. While for the 50 MPa mixtures a decrease in the drying shrinkage strain values between 28 and 53% for CWP replacement levels above 20% are observed.
Mixture I.D. | Cement | CWP | Fine aggregate | Coarse aggregate | Water content | Initial slump (mm) |
---|---|---|---|---|---|---|
M25-0 | 310 | 0 | 749 | 1102 | 190 | 110 |
M25-10 | 279 | 31 | 737 | 1105 | 190 | 130 |
M25-20 | 248 | 62 | 734 | 1101 | 190 | 103 |
M25-30 | 217 | 93 | 731 | 1097 | 190 | 95 |
M25-40 | 186 | 124 | 629 | 1093 | 190 | 55 |
M50-0 | 485 | 0 | 662 | 993 | 208 | 55 |
M50-10 | 437 | 48 | 658 | 988 | 208 | 65 |
M50-20 | 388 | 97 | 654 | 981 | 208 | 60 |
M50-30 | 340 | 145 | 650 | 975 | 208 | 42 |
M50-40 | 291 | 194 | 673 | 968 | 208 | 10 |
Mixtures’ proportions (kg/m3) and initial slump values (mm) (modified from [43]).
Mixture | Shrinkage strain (microstrain) | Mixture | Shrinkage strain (microstrain) |
---|---|---|---|
M25-0 | 2608 | M50-0 | 2569 |
M25-10 | 2488 | M50-10 | 2222 |
M25-20 | 2817 | M50-20 | 2413 |
M25-30 | 1033 | M50-30 | 1199 |
M25-40 | 1859 | M50-40 | 1848 |
Drying shrinkage strain values at 120 days (microstrain) (modified from [43]).
The concrete durability concerning its resistance to chloride ion penetration and chloride induced corrosion can be judged by the RCPT. The inclusion of CWP as partial cement replacement has a significant effect on the chloride ion penetration of the 25 and 50 MPa concrete mixtures. Figure 8 demonstrates a significant reduction in the 28 and 90 days’ test results of all CWP concrete mixtures. The COV ranged from 3 to 15%.
Chloride ion penetration.
At 28 days of age, the use of 20, 30 and 40% CWP reduces the total passed charge by 2–8 times lower than that of the control mixture. Mixtures with 30 and 40% are rated as “Very Low” for chloride ion penetration as per the classification of the ASTM C1202 [52]. At 90 days of age, the chloride ion penetration classification of all the 25 MPa mixtures including CWP is “Very low.” The reduction in the total passed charge for the mixtures incorporating CWP compared to its corresponding 28 days values ranged from 56 to 84%.
While for the 50 MPa mixtures, the 28 days chloride ion penetration decreases with the inclusion of CWP. The reduction is proportional to the CWP content. The reduction with respect to the control mixture is 38% for the use of 10% CWP and 90% for the use of 40% CWP. The ASTM classification of mixtures including high levels of CWP (i.e., ≥20) is shifted from “High” to “Low” and even “Very Low.” At the 90 days of age, chloride ion penetration for all 50 MPa CWP mixtures is classified as “Very Low.” This significant reduction could be due to the microstructure densification and refinement of the pore structure provided by the fine particles of CWP in addition to its pozzolanic effect. Also, the reduction with age indicates the development of a dense microstructure, especially with discontinuous pore system. Similar findings were reported in other studies [6, 30, 34, 56].
The corrosion protection of the concrete to the embedded reinforcement can be assessed by its electrical resistivity [57]. Figure 9 displays the bulk electrical resistivity at 28 and 90 days of age. The COV ranged from 4 to 10%. It should be noted that electrical resistivity is mainly affected by the porosity and the pore size distribution [58]. Therefore, the development of the microstructure could be judged by measuring the electrical resistivity. Ionic mobility is reduced by the discontinuity of pores, and hence concrete resistivity and corrosion protection will increase. The resistivity results of all concrete mixtures including CWP are higher than those of the control mixtures. Microfilling effect and pozzolanic activity of the CWP which could lead to a denser microstructure could be the main reasons for the increase in the resistivity of the mixtures including CWP. It was reported that the use of ceramic polishing residues was reported to reduce water permeability of cement mortar samples [6, 34].
Bulk electrical resistivity.
At 28 days of age, 25 MPa mixtures including 20, 30 and 40% CWP have a resistivity higher than 10 kΩ.cm. This is classified as “High” to “Very High” corrosion protection levels according to ACI 222R-01 [57]. The increase in resistivity is proportional to the CWP replacement level. At 90 days of age, using CWP demonstrates a significant increase in the electrical resistivity values with respect to the control mixture. The 50 MPa concrete mixtures with CWP had similar performance to the 25 MPa mixtures at both ages. Including 10% CWP results in a “High” corrosion protection level. When CWP is included with 20% or more the corrosion protection level is “Very High” at both ages.
Both RCPT and resistivity results confirm the performance of the concrete mixtures including CWP with regards to chloride ion attack, chloride-induced corrosion, and corrosion protection.
The permeable pores of the concrete mixtures can assess the development of the pore system and judge the microstructure development. Figure 10 shows the permeable pores measured at 90 days of age. The COV ranged from 2 to 8%. In general, the permeable pores are decreased by the inclusion of CWP compared to the control mixture.
Ninety days permeable pores.
In the case of the 25 MPa mixtures, the permeable pores are reduced by 17% up to 36% due to the inclusion of CWP as a partial cement replacement. Similar performance is observed for the 50 MPa mixtures. The reduction in pores volume ranged from 2 to 24% compared to the control mixture. The inclusion of the fine CWP particles with high SSA could physically have a microfilling effect and improves the particles’ packing in the mixtures. Also, to the CWP pozzolanic activity, the mixtures microstructure is densified. Therefore, the pore structure is refined resulting in lower pore volume. The reduction in permeable pores reduces the mobility of water from inside the concrete which is reflected in reducing the reduction in the drying shrinkage strain. Also, reduction in chloride ion penetration and immobility of ions are direct effects of the pores’ size refinement. This is reflected in the reduction of the chloride ions penetration and the improvement of the electrical resistivity with age.
MIP is a widely used test to characterize the pore structure of cement-based materials. The test is capable of providing information about the total porosity, and the median pore diameter based on intruded volume. The concrete pore system indicates its microstructural development that can be related to its performance.
Table 5 gives the results of the MIP test regarding total porosity percentage and the median pore diameter based on intruded volume at 90 days of age. The inclusion of CWP reduces the total porosity at 90 days of age. The use of 40% CWP as partial replacement of the cement reduces the porosity by 9 and 19% for the 25 and 50 MPa mixtures respectively compared to the same mixtures without CWP. The median pore diameter is reduced due to the inclusion of CWP. It is noted that the reduction was proportional to the CWP content. The reduction in the total porosity and the median pore diameter confirms the densification of the microstructure due to the inclusion of CWP as a partial cement replacement.
Mixture | Porosity (%) | Median pore diameter* (μm) |
---|---|---|
M25-0 | 21.297 | 4.2586 |
M25-10 | 20.015 | 4.0115 |
M25-20 | 19.754 | 3.7404 |
M25-30 | 19.135 | 3.6184 |
M25-40 | 19.437 | 3.4737 |
M50-0 | 22.426 | 4.0380 |
M50-10 | 21.131 | 3.8382 |
M50-20 | 19.415 | 3.5876 |
M50-30 | 18.944 | 3.5747 |
M50-40 | 18.126 | 3.4000 |
MIP results at 90 days of age.
Based on the intruded volume.
The reduction in the total porosity and especially the reduction in the pore size confirm the superior durability performance of the mixture observed at the late age. The microstructure development could be related to the durability performance. The median pore diameter was correlated to the 90 days RCPT and electrical resistivity values as shown in Figure 11. The median pore diameter correlates well with the durability test results. The correlation coefficient (R2) is 0.9517 and 0.7977 for the median pore diameter relationship with the RCPT and the electrical resistivity respectively.
Relation between median pore diameter and 90 days RCPT and electrical resistivity.
To better understand the performance of CVC mixtures including CWP, the main microstructural characteristics are inspected by scanning electron microscope (SEM). Microstructure examination is conducted at 90 days of age. The examination is conducted on the control mixture for both concrete grades (i.e., M25-0 and M50-0), and the mixtures including the highest CWP content (i.e., M25-40 and M50-40).
Figure 12 shows the SEM images of the general characteristics for M25-0 and M25-40. For the M25-0 mixture, crystalline hydration products are observed in addition to several pores. For M25-40, fewer pores with smaller size are noticed which indicates the densification of the microstructure that confirms the superior durability performance. Few crystalline hydration products are observed. Figure 13 displays the aggregate matrix interfacial transition zone (ITZ) for M25-0 and M25-40 mixtures. Crystalline hydration products are noticed in both mixtures in the ITZ region with smaller crystal size in M25-40 mixture. The matrix around the aggregate in the M25-40 mixture includes lesser pores compared to M25-0, this is similar to the observations of the general matrix microstructure.
SEM image of general microstructure for M25-0 and M25-40 mixtures.
SEM image of ITZ region for M25-0 and M25-40 mixtures.
The general microstructure for M50-0 and M50-40 is shown in Figure 14. Generally, the 50 MPa mixtures have a denser microstructure compared to the 25 MPa mixtures. For the M50-0 mixture, few pores are noticed, and the crystalline hydration products are smaller in size. The inclusion of CWP densified the microstructure by refining the pore structure as depicted in the SEM image. The ITZ region microstructure is presented in Figure 15. The incorporation of CWP improves the densification of the ITZ region microstructure. The crystalline hydration products and pores’ size are reduced due to the inclusion of CWP.
SEM image of general microstructure for M50-0 and M50-40 mixtures.
SEM image of ITZ region for M50-0 and M50-40 mixtures.
Self-compacting concrete (SCC) has received wide attention and used in the construction industry worldwide since its development [59]. SCC is featured with high fluidity, and at the same time, high resistance to segregation and is placed purely under its weight without the need for vibration [60, 61, 62]. SCC properties are the result of modifying the composition of CVC by incorporating high powder content that has been mainly cement. However, the use of high cement content is not desirable as it will increase the cost and has other negative environmental effects. Replacing cement in SCC mixtures with waste powder is a trend gaining a great deal of attention with the growing awareness toward environmental protection and sustainable construction [63, 64, 65, 66, 67, 68, 69, 70]. CWP is used to partially replace cement to produce eco-friendly SCC. The cement content in the control mixture is 500 kg/m3 based on the preliminary mix design. The powder content of the control mixture meets the recommended value by EFNARC specifications [71]. The cement is partially replaced by the CWP in 20, 40 and 60% by weight. The concrete mixture is expected to yield compressive strength in the range of 80 MPa. The details of the mixtures’ proportions are given in Table 6.
Ordinary Portland cement (OPC) is used as the main binder. The specific surface area of cement is 380 m2/kg. Natural crushed stone of maximum size 9.5 mm is used as coarse aggregate. The specific gravity is 2.65 while the absorption was 0.7%. Natural sand with fineness modulus between 2.5 and 2.7 is used as fine aggregate. The specific gravity is 2.63.
Several tests are conducted to investigate the effect of replacing cement with CWP on the fresh properties of the produced concrete. Unconfined flowability of the produced SCC mixture is assessed by the slump flow test in accordance to ASTM C1611 [72]. Passing ability is evaluated through two tests namely the J-ring (i.e., ASTM C1621 [73]), and L-box. The segregation resistance is measured through conducting the GTM segregation column test conforming to ASTM C1610 [74]. Finally, the viscosity is measured by following the V-funnel test procedure described in the EFNARC specification [71]. On the other hand, compressive strength is performed at two test ages (i.e., 7 and 28 days) in order to evaluate the strength development. The durability characteristic is evaluated by conducting the bulk electrical resistivity as per ASTM C1760 [53] at 28 and 90 days of age. Triplicate samples are used to conduct the compressive strength and the bulk electrical resistivity tests and the average results are used. Figure 16 shows the different tests conducted. The microstructure development is judged by measuring the permeable pore volume at 28 and 90 days of age. Also, the pore system (i.e., total porosity and median pore diameter) is assessed using mercury intrusion porosimetry (MIP). The MIP is conducted at 90 days of age.
Different tests conducted on SCC.
Slump flow test evaluated the unconfined flowability of the produced SCC mixtures. Figure 17 displays the test results together with the EFNARC specifications [71].
Slump flow results.
It is noticed that the slump flow decreases as the amount of CWP in the mixture increases. Even with the reduction in the slump flow values, none of the CWP mixtures dropped to the slump flow class one (SF1) which is critical in the presence of highly congested reinforced concrete structures.
Chopra and Siddique [48] reported a similar trend when using rice husk ash (RHA) as cement replacement. The relatively higher specific surface area (SSA) of the CWP compared with cement would increase the water demand and accordingly resulted in lower slump flow values. Similarly, Sfikas et al. [75] reported a reduction in the slump flow of SCC when they used metakaolin, which is characterized by a high SSA, to replace cement.
The time taken for concrete to reach the 500 mm diameter circle on the steel base plate of the slump flow test is measured (T50). The T50 value can judge the viscosity of the SCC mixtures. High T50 values indicate mixtures with higher viscosity. The T50 results are given in Table 7.
Mixture ingredients | Mixture designation | |||
---|---|---|---|---|
Control | R-20 | R-40 | R-60 | |
Cement | 500 | 400 | 300 | 200 |
CWP | 0 | 100 | 200 | 300 |
Water | 175 | 175 | 175 | 175 |
Fine aggregate | 871 | 871 | 871 | 871 |
Coarse aggregate | 871 | 871 | 871 | 871 |
Super plasticizer | 8.33 | 8.72 | 8.33 | 8.80 |
VMA* | 1.6 | 1.6 | 1.6 | 1.6 |
w/cm** | 0.35 | 0.35 | 0.35 | 0.35 |
Mixtures’ proportions for SCC (kg/m3).
VMA = viscosity-modifying admixture.
w/cm = water/(cement + slag or CWP).
Property measured | Control | R-20 | R-40 | R-60 |
---|---|---|---|---|
T50 (seconds) | 2.68 | 2.47 | 3.24 | 4.04 |
V-Funnel (seconds) | 10.4 | 10.01 | 11 | 12.82 |
L-box ratio (H2/H1) | 0.963 | 0.966 | 0.977 | 0.967 |
Fresh test results (modified from [42]).
The passing ability of SCC is evaluated by the J-ring test. This test evaluates how the SCC mixtures can perform in the presence of reinforcing bars in form works. The difference between the unrestricted slump flow diameter and the J-ring flow diameter is shown in Figure 18. The inclusion of CWP improves the passing ability of the SCC mixtures. As the CWP content increases the mixtures’ passing ability is improved and shows a great capacity for flowing through congested spaces. Therefore, mixtures containing high CWP perform better than the control mixture with regards to the passing ability.
J-ring results.
The passing ability of SCC through congested reinforcement can also be assessed by using the L-box test. The L-box results are given in Table 7. Comparable blocking ratios are observed for all tested mixtures. The variation is less than 1.5%. SCC mixtures including CWP mixtures show no signs of blocking. Generally, EFNARC [71] suggests blocking risk is likely if the blocking ratio is below 0.8. The viscosity of the mixtures is too high if the blocking ratio is less than 0.8. This can cause blocking around highly congested sections. Based on the results, all mixtures with CWP can be used in applications where flow through congested reinforcement is needed.
In this test, the viscosity and filling ability of the fresh concrete is judged by the V-funnel test where the concrete is forced to flow through small cross sections and confined spaces. The flow rate (i.e., V-funnel time) of the SCC through the small cross-section is directly related to the mixture’s viscosity.
The V-funnel test results are given in Table 7. The V-funnel results show an increasing trend, indicating a higher viscosity of the mixtures. All the measured V-funnel time values correspond to the second viscosity class according to EFNARC specification [71]. The increase in the viscosity indicates an improvement in the segregation resistance. The viscosity-modifying admixture (VMA) is typically used to adjust mixtures’ viscosity and enhance segregation resistance. Since the mixtures’ viscosity values are significantly enhanced by the incorporation of CWP the VMA could be eliminated from the mixture or its dosage could be reduced. This would result in more economical and low-cost mixtures.
The ability of concrete to remain homogeneous in the composition in its fresh state is defined as its segregation resistance. The GTM segregation column test is used to evaluate the mixtures’ segregation resistance.
Segregation percentage is shown in Figure 19. The segregation percentage decreases as the CWP content increases in the mixtures. The CWP significantly improves the segregation resistance of the SCC mixtures. The incorporation of CWP in SCC enhances the cohesiveness characteristics of the mixtures. The segregation percentages are below 15%, which shows that the SCC mixtures were superior regarding segregation resistance. Segregation resistance is related to viscosity. The improvement in segregation resistance is confirmed by the V-funnel test results. As the amount of CWP increases in the mixtures from 0 to 60%, the segregation resistance is enhanced by 72.5%. The substantial enhancement in the segregation resistance can be explained by the fact that the water adsorption of the CWP particles may induce suction forces possibly leading to cluster formation. This will lead to an increase in the inter-particle bonds as in the clustering theory enhancing the segregation resistance similar to RHA mixtures studied by Le and Ludwig [76].
Segregation resistance results.
Strength is measured at different test ages (7, 28, and 90 days) to evaluate the strength development as affected by the inclusion of CWP as partial cement replacement. The strength development due to the inclusion of any cement replacing material is mainly affected by the cement hydration and pozzolanic reaction the used material, and the effect on the concrete microstructure especially the densification of the microstructure with a particular focus on the aggregate-paste interfacial zone [77].
Figure 20 shows the compressive strength development with age. The COV ranged from 0.4 to 3.0%. At the 7 days of age, it is noticed that the inclusion of CWP decreases the strength and the reduction is proportional to the CWP content. This could be a direct result of replacing cement by CWP which has no hydraulic reaction. At the 28 days of age, the mixture including 20% by weight CWP showed higher strength compared to the control mixture. Nevertheless, the mixture of 60% by weight CWP shows the least developed strength. Since CWP is characterized by the slow pozzolanic reaction, it is expected not to see much effect until late ages. At the 90 days of age, the improvement in strength is noticeable. At the 90 days of age, mixtures with 20 and 40% by weight CWP achieve the highest compressive strength compared to the control mixture. This implies that 20–40% by weight CWP is the optimum cement replacement to obtain high compressive strength.
Compressive strength development with age.
The increase in the strength can also be explained through the nucleation sites (i.e., nucleation of CH around the CWP particles). The CWP improves the packing of the concrete mixture due to its high SSA and its pozzolanic reaction, and the cement hydration acceleration similar to the effect of rice husk ash (RHA) observed in another investigation [76]. On the other hand, the use of 60% by weight CWP shows marginal improvement in strength; this can be due to the high amount of silica from the CWP, and the insufficient amount of calcium hydroxide (CH) from the cement hydration. Hence, some silica is left without chemical reaction. Similar behavior was observed by using RHA (i.e., characterized by high SSA and high silica content) as cement replacement [48].
The electrical resistivity of concrete is affected by several factors such as porosity, pore size distribution, connectivity, concrete’s moisture content, and ionic mobility in pore solution. Electrical resistivity assesses the concrete protection of reinforcing steel against corrosion. According to ACI 222R-01 [57], the corrosion protection level is improved as the resistivity value increases.
The resistivity values are presented in Figure 21 at 28 and 90 days of age. The COV ranged from 6.4 to 13.2%. The resistivity increases with age. The inclusion of CWP significantly increases the mixtures’ resistivity. The significant increase in the resistivity due to the inclusion of CWP suggests that CWP tended to reduce the interconnected pore network contributing to the reduction of the concrete’s conductivity. With age, CWP pozzolanic activity contributes to the refinement of concrete pores and microstructure, thus further reduces the ionic mobility and hence the concrete’s conductivity. The improved resistivity indicated that the durability of the CWP concrete mixtures to protect reinforcing steel against the corrosive environment is much better than that of the control mixture without CWP.
Electrical resistivity of SCC.
The MIP test provides information about the pore system (i.e., pore volume and median pore diameter). The MIP results can help understand the development of the concrete microstructure and can also explain the other obtained results. Table 8 shows the MIP test results at 90 days of age. Test results show that high CWP content has a significant reduction of the pore volume and the pores’ size. The reduction in the pore volume and the pores’ size indicates densification of the microstructure. Also, the MIP results confirm the improvement observed in the resistivity results and compressive strength.
Zero-cement alkali-activated concrete (AAC) emerged as an alternative to cement-based concrete [78, 79, 80, 81, 82, 83, 84]. Sometimes, AAC is referred to as inorganic-polymer or geopolymer concrete. In AAC, cement is completely replaced. AAC utilizes and silica and alumina rich materials to be alkali-activated to form a three-dimensional CaO-free alumino-silicate binder. AAC offers a significant opportunity for the reuse of several industrial by-products and wastes such as fly ash, metakaolin, and blast-furnace slag. Geopolymerization technology is based on the reaction of alkaline solutions such as sodium hydroxide (NaOH), potassium hydroxide (KOH) and sodium silicate solution. The CWP is characterized by its high silica and alumina content which makes it a good candidate to be used in making ACC. The limited studies on suing CWP in AAC [38, 39, 40] concluded that the optimum curing temperature ranges from 60 to 80°C, the curing period ranges between 24 and 48 hours, and the molarity of the alkali solution is 12 M.
The use of CWP in the making AAC still needs further investigations to develop a better understanding of its performance. CWP is used to make AAC using different alkali solutions, mainly NaOH and KOH. Several parameters are investigated which include alkaline solutions with 12 M concentration (i.e., NaOH alone, KOH alone and combination), CWP to aggregate ratio (i.e., 1:1.5–1:2.0–1:2.5), admixture dosage (i.e., 1.5 and 4.0%), curing time (i.e., 60°C for 24 and 48 hours), the inclusion of slag in addition to CWP (i.e., slag content 10, 20 and 40%). Several tests are used to evaluate the performance of the mixtures which include flowability (i.e., ASTM C1437 [85]), cube compressive strength, permeable pores (i.e., ASTM C642 [54]), initial rate of water absorption (i.e., ASTM C1585 [86]), and electrical resistivity (i.e., ASTM C1760 [53]). The COV ranged from 0.3 to 2.8%.
The sodium hydroxide flakes and potassium hydroxide are dissolved in distilled water to make a solution with the desired concentration (i.e., 12 M) at least 1 day before its use. Table 9 shows the alkali solutions used and the combination of NaOH and KOH solutions. The dry ingredients are first mixed for about 1 minute. The sodium hydroxide and potassium hydroxide solutions are added to the dry materials based on the order of mixing in Table 9 and mixed for 3 minutes.
The effect of aggregate content was evaluated by the flowability and 7 days compressive strength. Mixtures are cured at 60°C for 24 hours. Figure 22 shows the flowability and 7 days compressive strength as affected by the CWP to aggregate ratio. It is noticed that the flowability decreases as the aggregate content increases. This is similar to the behavior cement concrete as the CWP content acts as a lubricant between aggregate particles. Oppositely the 7 days compressive strength improved by the increase of the aggregate content. The mixing regime of the solution affects the flowability and strength. The mixing regime (A) shows the best flowability performance while the other mixing regimes show similar flowability values. The mixing regimes (D) and (E) produce the highest compressive.
Flowability and 7 days compressive strength as affected by CWP to aggregate ratio.
Superplasticizer (i.e., polycarboxylic ether based) is added with a dosage of 1.5 and 4.0% of the CWP weight. The AAC mixture with CWP to the aggregate ratio (1:2.5) and 24 hours curing at 60°C is used to examine the effect of admixture dosage. Flowability and the 7 days compressive strength results are presented in Table 10. The use of 1.5% by weight superplasticizer, shows variable improvement in the flowability and marginal improvement in the strength. By increasing the admixture dosage to 4.0%, the flowability and strength are improved. For both admixture dosages, the mixing regimes (D) and (E) show the best flowability improvement and highest compressive strength.
I.D. | Alkali solutions % | Mixing regime of the solutions with the CWP | |
---|---|---|---|
KOH | NaOH | ||
A | 0 | 100 | — |
B | 100 | 0 | — |
C | 20 | 80 | NaOH solution is added first and mixed with solids for 1 minute, then KOH is added and mixing continues for an additional 2 minutes |
D | 40 | 60 | NaOH and KOH solutions are mixed then added to solids and mixed for 3 minutes |
E | 60 | 40 | KOH solution is added first and mixed with solids for 1 minute, then NaOH is added and mixing continues for an additional 2 minutes |
Mixtures’ I.D., alkali solutions used and mixing regime of solutions.
The AAC mixture with CWP to aggregate ratio (1:2.5) and 4% admixture is used to examine the effect of curing time (i.e., 24 and 48 hours) at 60°C. Figure 23 shows the effect of curing time on the 7 days compressive strength. The compressive strength increases as the curing time increases. A similar trend is reported for metakaolin-based AAC [87]. Although increasing the curing time improves the compressive strength, the application of shorter curing time is considered from the point of reducing the energy consumption.
Seven days compressive strength for the AAC mixture with CWP to aggregate ratio 1:2.5 as affected by curing time at 60°C.
Several studies investigated the use of slag in making AAC [88, 89, 90, 91, 92]. Slag proved to be a suitable material in making AAC. Slag is characterized by having some hydraulic reaction due to the existence of calcium oxide (CaO) beside the existence of silica and alumina for the alkali activation. Therefore, slag is used to replace part of the CWP. This will help improve the flowability of the AAC mixture and improve the strength development without the need to increase curing time. The AAC mixture with CWP to aggregate ration 1:2.5 and 4% admixture is used to assess the effect of including slag as a binder material in addition to the CWP. The slag replaced the CWP with 10, 20 and 40% by weight. The AAC mixtures including slag are subjected to three curing regimes; air curing, 24 hours at 60°C followed by air curing, and 24 hours at 60°C followed by water curing for 6 days. Figure 24 shows the flowability of AAC mixtures including slag and CWP. The inclusion of slag improves the mixtures’ flowability. The improvement is proportional to the slag content with the highest improvement at 40% slag.
Flowability of AAC including CWP and slag.
The effect of including slag with CWP on the 7 days strength is displayed in Figure 25. The air cured mixtures showed the lowest strength development. It is observed that the (oven + air) and (oven + water) results are comparable for both the 20 and 40% slag replacements. The strength values are found to increase with the increase in slag % replacing the CWP, with the highest at 40% slag.
Seven days compressive strength of AAC including CWP and slag.
The inclusion of slag is beneficial in producing AAC using CWP with a level of replacement of 40%. Based on the flowability and the 7 days compressive strength, the following are the optimum mixture’s parameter to make AAC using CWP:
the CWP to the aggregate ratio is 2.5,
the alkali solutions mixing regime (D) (i.e., NaOH 60% and KOH 40% mixed) produces suitable flowability and strength;
the use of 4% of superplasticizer to improve flowability;
the application of 24 hours at 60°C followed by air curing; and
the use of 40% by weight slag to replace CWP.
The performance of an AAC mixture following the above parameters is assessed. Table 11 summarizes the obtained results. Results show that CWP in combination with 40% slag can produce AAC with strength suitable for different structural applications. The electrical resistivity and initial rate of absorption indicate that the produced AAC is characterized by high durability. The change in the test results values with age indicates that most of the reactions are finished at 7 days of age. Hence there is no need for waiting to evaluate the performance at 28 days of age similar to Portland cement concrete.
Test age (days) | ||
---|---|---|
7 | 28 | |
Compressive strength (MPa) | 39.3 | 40.7 |
Permeable pores % | 8.89 | 8.32 |
Electrical bulk resistivity (kΩ.cm) | 17.9 | 18.2 |
Initial rate of absorption (mm/min1/2) sorptivity | 0.15 | 0.12 |
Seven and twenty-eight days results for optimum AAC mixture.
The CWP contains high silica and alumina content (i.e., >80%). Also, it is characterized by having some amorphous content which shows pozzolanic activity especially at late ages. Therefore, CWP has strong potentials to be used as an ingredient in making eco-friendly concretes.
Using CWP as an ingredient in making CVC is viable. High-performance concrete can be produced by including CWP as partial cement replacement. CWP improves the workability retention of the CVC mixtures. The inclusion of CWP will reduce the early-age strength and slowed the strength development. Significant improvement of CVC durability can be achieved by including high content of CWP. The CVC performance varies according to the CWP content. CWP can be used in the range of 10–20% to improve workability retention and late strength development. A CWP content ranging from 30 to 40% is needed to improve durability. If the performance of mixture requires the combination of workability retention, strength and durability, a CWP content ranging from 20 to 30% can be used to optimize all required characteristics.
CWP can be used as a partial cement replacement to produce SCC that meets international requirements. All fresh concrete properties, except for slump flow, are significantly improved by the incorporation of CWP. The improvement is proportional to the CWP content. Similar to CVC, the inclusion of CWP affected the strength development and enhanced the durability. SCC with improved fresh performance and optimized strength can be produced using 40% CWP as partial cement replacement.
The use of CWP in making AAC showed promising potentials. The production of AAC using CWP should consider the aggregate content of the mixture, the use of superplasticizer admixtures and the use of an alkali solution composed of NaOH and KOH. The combination of slag with CWP improves the workability and strength development without the need for long curing time to conserve energy. The combination of CWP with fly ash can also be an alternative to enhance the performance of the produced AAC.
Finally, CWP has encouraging potentials to be used as an ingredient to make eco-friendly conventional-vibrated concrete (CVC), self-compacting concrete (SCC) and zero-cement alkali-activated concrete (AAC). The concrete industry can and will play a vital role in the sustainable development through the utilization of industrial waste materials.
This work was financially supported by the UAEU-UPAR2 Research Grant # 31 N2018. Also, the donation of the ceramic waste powder for the study by PORCELLAN (ICAD II MUSSAFAH—ABU DHABI, UAE) is much valued. The help of master students Dima M. Kanaan and Sama T. Aly is highly appreciated. Support to the second author by Southern Plains Transportation Centre (SPTC) to University of New Mexico is much appreciated.
The Edited Volume, also known as the IntechOpen Book, is an IntechOpen pioneered publishing product. Edited Volumes make up the core of our business - and as pioneers and developers of this Open Access book publishing format, we have helped change the way scholars and scientists publish their scientific papers - as scientific chapters.
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\n\nCan collaboration be inspired by a publishing format? At IntechOpen, the answer is yes. The way the research is published, the way it is accessed, it’s all part of our mission to help academics make a greater impact by giving readers free access to all published work.
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\n\nCURRENT PROJECTS
\n\nTo view current Open Access book projects that are Open for Submissions visit us here.
\n\nNot sure if this is the right publishing option for you? Feel free to contact us at book.department@intechopen.com.
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