A ceramic fuel cell in an all solid-state energy conversion device that produces electricity by electrochemically combining fuel and oxidant gases across an ionic conducting oxide. Current ceramic fuel cells use an oxygen-ion conductor or a proton conductor as the electrolyte and operate at high temperatures (>600°C). Ceramic fuel cells, commonly referred to as solid-oxide fuel cells (SOFCs), are presently under development for a variety of power generation applications. This paper reviews the science and technology of ceramic fuel cells and discusses the critical issues posed by the development of this type of fuel cell. The emphasis is given to the discussion of component materials (especially, ZrO2 electrolyte, nickel/ZrO2 cermet anode, LaMnO3 cathode, and LaCrO3 interconnect), gas reactions at the electrodes, stack designs, and processing techniques used in the fabrication of required ceramic structures.

Ceramic Fuel Cells

Materials Challenges in Nuclear Energy

Despite being first demonstrated over 160 years ago, and offering significant environmental benefits and high electrical efficiency, it is only in the last two decades that fuel cells have offered a realistic prospect of being commercially viable. The solid oxide fuel cell (SOFC) offers great promise and is presently the subject of intense research activity. Unlike other fuel cells the SOFC is a solid-state device which operates at elevated temperatures. This review discusses the particular issues facing the development of a high temperature solid-state fuel cell and the inorganic materials currently used and under investigation for such cells, together with the problems associated with operating SOFCs on practical hydrocarbon fuels.

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Solid oxide fuel cells

Progress in material selection for solid oxide fuel cell technology: A review

Present review is aimed at providing a state-of-art development of anode for solid oxide fuel cell (SOFC) with principal emphasis on the materials aspect. The criteria for the anode of SOFC are first presented. The prospects and problems of the currently developed anode materials are elucidated. In particular, the electrochemical properties of the Ni/YSZ cermet anode that is the most commonly employed in the establishment of SOFC stack is described along with various approaches attempted for their improvements. The advantages and disadvantages of other anode materials are compared to offer some insights for the research and development of new generation of anode materials for SOFC.

A review on the status of anode materials for solid oxide fuel cells

The problem of evaluating structural integrity of primary water reactor steam generator tubes during normal operation, main steam line break, and severe accidents is addressed Failure models for intact tubes as well as tubes with axial and circumferential machined flaws are discussed Two types of tube failure are identified—ligament rupture and unstable burst The models are generalized for application to complex stress corrosion cracks that are observed in the field Results from rupture tests of flawed steam generator tubes are discussed and used to validate the failure models

Structural integrity assessment of nuclear steam generator

Fracture and deformation problems anticipated for the first wall of a future steady-state fusion power reactor are discussed. The stress sources considered are surface heat flux, irradiation-induced creep and swelling, and coolant pressure. Simple examples, such as a beam or a plate, as well as finite element analyses of more realistic structures from recent design studies, are used to explain the problems that are anticipated. Detailed two- and three-dimensional finite element analyses of a grooved first wall are presented to show that grooving the first wall is a potential solution for high thermal stress.

Stress-related problems in tokamak fusion reactor first walls

This report provides an update of an earlier assessment of environmentally assisted fatigue for components in light water reactors. This report is a deliverable in September 2016 under the work package for environmentally assisted fatigue under DOE’s Light Water Reactor Sustainability program. In an April 2016 report, we presented a detailed thermal-mechanical stress analysis model for simulating the stress-strain state of a reactor pressure vessel and its nozzles under grid-load-following conditions. In this report, we provide stress-controlled fatigue test data for 508 LAS base metal alloy under different loading amplitudes (constant, variable, and random grid-load-following) and environmental conditions (in air or pressurized water reactor coolant water at 300°C). Also presented is a cyclic plasticity-based analytical model that can simultaneously capture the amplitude and time dependency of the component behavior under fatigue loading. Results related to both amplitude-dependent and amplitude-independent parameters are presented. The validation results for the analytical/mechanistic model are discussed. This report provides guidance for estimating time-dependent, amplitude-independent parameters related to material behavior under different service conditions. The developed mechanistic models and the reported material parameters can be used to conduct more accurate fatigue and ratcheting evaluation of reactor components.

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Study the Cyclic Plasticity Behavior of 508 LAS under Constant, Variable and Grid-Load-Following Loading Cycles for Fatigue Evaluation of PWR Components

A solid-breeder, water-cooled blanket option based on a multilayer configuration was developed for the ITER (International Thermonuclear Experimental Reactor). The blanket uses beryllium for neutron multiplication and lithium oxide for tritium breeding. The material forms are sintered products for both materials with 0.8 density factor. The lithium-6 enrichment is 90%. The blanket can accommodate a factor of two change in the neutron wall loading without violating the different design guidelines. The design philosophy adopted for the blanket is to produce the necessary tritium required for the ITER operation and to operate at power reactor conditions as much as possible. At the same time, the reliability and the safety aspects of the blanket are enhanced by using a low-pressure coolant and separating the tritium purge lines from the coolant system. The blanket modules are made by hot vacuum forming and diffusion bonding a double-wall structure with integral cooling channels. The different aspects of the blanket design including tritium breeding, nuclear heat deposition, activation analyses, thermal-hydraulics, tritium inventory, structural analyses, and water-coolant conditions are summarized. >

/pdf/solid-breeder-blanket-option-for-the-iter-conceptual-design-567929o8jj.pdf

Saurin Majumdar

Papers

Structural integrity assessment of nuclear steam generator

Stress-related problems in tokamak fusion reactor first walls

Study the Cyclic Plasticity Behavior of 508 LAS under Constant, Variable and Grid-Load-Following Loading Cycles for Fatigue Evaluation of PWR Components

Solid breeder blanket option for the ITER conceptual design

Analysis of pushout tests on an SiC-fiber-reinforced reaction-bonded Si3N4 composite☆