Assessment of Silicon Carbide Composites for Advanced Salt-Cooled Reactors

Katoh, Yutai; Wilson, Daniel A.; Forsberg, Charles W.

doi:10.2172/982717

Cited by 21 publications

(19 citation statements)

References 134 publications

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“…Hexagonal replaceable graphite elements are placed immediately adjacent to the fuel assemblies, and a permanent graphite reflector is located around the outer perimeter of the core. The beryllium in the primary salt provides some additional moderation, and the lithium is isotopically pure 7 Li to minimize tritium production. The upper core support plate holds the core components in place and restrains the core components against the upward salt flow.…”

Section: Fig 2 Cross Section Of a Fuel Platementioning

confidence: 99%

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AHTR Mechanical, Structural, and Neutronic Preconceptual Design

Varma¹,

Holcomb²,

Peretz³

et al. 2012

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Section: Fig 2 Cross Section Of a Fuel Platementioning

confidence: 99%

“…6  Initial advanced materials options assessments have been performed. [7][8][9]  Component demonstration planning has been initiated. 10 …”

Section: Introductionmentioning

confidence: 99%

AHTR Mechanical, Structural, and Neutronic Preconceptual Design

Varma¹,

Holcomb²,

Peretz³

et al. 2012

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“…An excellent review of the performance of SiC composites for application in molten salt environments was performed by Katoh, Wilson, and Forsberg (2007). The report focused on the possible use of SiC for reactor components such as fuel element cladding, control rods, and intermediate heat exchangers.…”

Section: Mechanical Propertiesmentioning

confidence: 99%

Feasibility Study of Secondary Heat Exchanger Concepts for the Advanced High Temperature Reactor

Sabharwall¹

2011

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An evaluation of viable secondary heat exchanger concepts for use with a molten salt nuclear reactor is presented based on thermal design studies using KF-ZrF 4 as the secondary coolant. Along with the potential issues related to materials, fabrication methods, system configuration, coolant properties and hot corrosion, a discussion of challenges associated with integrating a steam power cycle with the reactor is also presented. The reference reactor design and configuration is a 3,400 MW t advanced high temperature reactor utilizing Rankine subcritical or supercritical cycles for power production. vi SUMMARYAdvanced High Temperature Reactors (AHTRs) will produce high outlet temperatures (704°C) using coated particle fuel, which provides for improved efficiency in power production or process heat for a variety of industrial applications. The AHTR is part of the fluoride-salt-cooled high temperature reactor (FHR) class of nuclear reactors included in the advanced reactor concept program. Early design decisions for the AHTR include use of an intermediate heat exchanger (IHX) to couple the primary loop to an intermediate loop. Heat from the IHX will be circulated in the intermediate loop through a secondary heat exchanger (SHX) and transferred to a power cycle that produces electricity or transferred to an industrial process. The feasibility of design options for the SHX is the subject of this study, and for the purposes of this study, the following preliminary decisions were made as a baseline for comparison:x The reactor outlet temperature is 704°C, from which the primary, secondary and tertiary fluid temperatures are derived. x The intermediate loop fluid is KF-ZrF 4 , which determines the SHX upstream side fluid and fluid conditions. x A supercritical or subcritical steam cycle is connected downstream of the SHX for power production, setting the fluid conditions on the downstream side. x The SHX is either a helically coiled tube and shell design or a printed circuit heat exchanger (PCHE) design. x The baseline material of construction is Alloy N.Alternative parameters are evaluated in this study and compared to the baseline based on the feasibility of their implementation.Heat exchangers are considered key components that need to be extensively investigated because they are operated under a severe environment and their performance is directly related to the overall system efficiency and safety. Previous analysis (INL/EXT-11-21584) of the power conversion system showed that Rankine subcritical and supercritical cycles with a turbine inlet temperature of 679°C can yield conversion efficiencies of 42% and 44% respectively, with KF-ZrF 4 as the secondary salt coolant.Fluoride salts such as KF-ZrF 4 have highly desirable characteristics for high temperature nuclear reactor heat transport. A characteristic of molten fluoride salts is that they can easily dissolve passive oxide layers such as the chromia oxide film (Cr 2 O 3 ) that provides the corrosion resistance for stainless steels and nickel based alloys...

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“…It should be noted that the stability under irradiation is a unique characterisitic of SiC‐SiC CMCs, which is not exhibited by other CMCs including those reinforced by carbon fibers. Besides application in LWRs, SiC‐SiC CMCs have potential structural and insulation applications in other nuclear energy systems, such as Very High Temperature Reactor (VHTR), Gas‐Cooled Fast Reactor (GFR), Molten Salt Reactor (MSR), Sodium Fast Reactor (SFR), and fusion reactors . Although SiC‐SiC CMC are manufactured in the forms of chopped fiber composites, particulate‐reinforced composites and continuous fiber composites, continuous fiber SiC‐SiC CMC are most suitable for fuel cladding fabrication.…”

Section: Introductionmentioning

confidence: 99%

Interlaboratory round robin study on axial tensile properties of SiC‐SiC CMC tubular test specimens

Singh

Gonczy²,

Deck

et al. 2018

Int J Applied Ceramic Tech

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An interlaboratory round robin study was conducted on the tensile strength of SiC‐SiC ceramic matrix composite (CMC) tubular test specimens at room temperature with the objective of expanding the database of mechanical properties of nuclear grade SiC‐SiC and establishing the precision and bias statement for standard test method ASTM C1773. The mechanical properties statistics from the round robin study and the precision statistics and precision statement are presented herein. The data show reasonable consistency across the laboratories, indicating that the current C1773‐13 ASTM standard is adequate for testing ceramic fiber reinforced ceramic matrix composite tubular test specimen. It was found that the distribution of ultimate tensile strength data was best described with a two‐parameter Weibull distribution, while a lognormal distribution provided a good description of the distribution of proportional limit stress data.

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Assessment of Silicon Carbide Composites for Advanced Salt-Cooled Reactors

Cited by 21 publications

References 134 publications

AHTR Mechanical, Structural, and Neutronic Preconceptual Design

AHTR Mechanical, Structural, and Neutronic Preconceptual Design

Feasibility Study of Secondary Heat Exchanger Concepts for the Advanced High Temperature Reactor

Interlaboratory round robin study on axial tensile properties of SiC‐SiC CMC tubular test specimens

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