TRISO-Coated Particle Fuel Performance

Petti, David A.; Demkowicz, Paul A.; Maki, John T.; Hobbins, R.R.

doi:10.1016/b978-0-08-056033-5.00055-0

Cited by 31 publications

(15 citation statements)

References 25 publications

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“…The TRISO particle fuel, which has been experienced and wellproven in the various High Temperature Gas Cooled Reactor (HTGR) programs and recently re-qualified by the DOE Next Generation Nuclear Plant (NGNP) program, shows excellent performance in radioactivity resistance capability through the presence of multiple layers of ceramic coating, which are chemically stable and mechanically strong at high temperature and high burnup conditions (Pappano et al, 2008;Petti et al, 2012).…”

Section: Introductionmentioning

confidence: 99%

Safety evaluation of accident-tolerant FCM fueled core with SiC-coated zircalloy cladding for design-basis-accidents and beyond DBAs

Chun

Lim

Chung

et al. 2015

Nuclear Engineering and Design

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h i g h l i g h t s• Thermal conductivity model of the FCM fuel was developed and adopted in the MARS.• Scoping analysis for candidate FCM FAs was performed to select feasible FA.• Preliminary safety criteria for FCM fuel and SiC/Zr cladding were set up.• Enhanced safety margin and accident tolerance for FCM-SiC/Zr core were demonstrated. Classification: L. safety and risk analysis a b s t r a c tThe FCM fueled cores proposed as an accident tolerant concept is assessed against the design-basisaccident (DBA) and the beyond-DBA (BDBA) scenarios using MARS code. A thermal conductivity model of FCM fuel is incorporated in the MARS code to take into account the effects of irradiation and temperature that was recently measured by ORNL. Preliminary analyses regarding the initial stored energy and accident tolerant performance were carried out for the scoping of various cladding material candidates.A 16 × 16 FA with SiC-coated Zircalloy cladding was selected as the feasible conceptual design through a preliminary scoping analysis. For a selected design, safety analyses for DBA and BDBA scenarios were performed to demonstrate the accident tolerance of the FCM fueled core. A loss of flow accident (LOFA) scenario was selected for a departure-from-nucleate-boiling (DNB) evaluation, and large-break loss of coolant accident (LBLOCA) scenario for peak cladding temperature (PCT) margin evaluation. A control element assembly (CEA) ejection accident scenario was selected for peak fuel enthalpy and temperature. Moreover, a station blackout (SBO) and LBLOCA without a safety injection (SI) scenario were selected as a BDBA. It was demonstrated that the DBA safety margin of the FCM core is satisfied and the time for operator actions for BDBA s is evaluated.

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Section: Introductionmentioning

confidence: 99%

Safety evaluation of accident-tolerant FCM fueled core with SiC-coated zircalloy cladding for design-basis-accidents and beyond DBAs

Chun

Lim

Chung

et al. 2015

Nuclear Engineering and Design

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“…For operational and economic reasons the preferred fuel kernel is low enriched uranium dioxide for the pebble bed reactors and uranium oxycarbide for the prismatic designs. TRISO fuel fabrication routes have been reviewed by Sawa and Petti et al . who describe TRISO‐coated particle fuel performance.…”

Section: Future Nuclear Applications Of Ceramicsmentioning

confidence: 99%

Opportunities for Advanced Ceramics and Composites in the Nuclear Sector

et al. 2013

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Ceramics have played a crucial role in the development of fission based nuclear power, in glass & glass composite high level wasteforms, in composite cements to encapsulate intermediate level wastes (ILW) and also for oxide nuclear fuels based on UO2 and PuO2/UO2 mixed oxides. They are also used as porous filters with the ability to absorb radionuclides (RN) from air and liquids and are playing a key role in the cleanup at Fukushima. Non‐oxides also find current fission applications including in graphite moderators and B4C control rods. Ceramics will continue to be significant in the near‐term expansion of nuclear power via next‐step developments of fuels with inert matrices or based on thoria and in wasteforms using alternative composite cements or single or multiphase ceramics that can host Pu & other difficult RN. Longer term advances for Generation IV reactors, which will operate at higher temperatures & with higher fuel burn‐up require innovative fuel developments potentially via carbides & nitrides or composite fuel systems. Novel non‐thermal (cement‐like) and thermal techniques are currently being developed to treat some of the difficult legacy wastes. Non‐thermally derived wasteforms developed from geopolymers, composite cements, hydroceramics, and phosphate‐bonded ceramics and thermally derived wasteforms made by Hot Isostatic Pressing and fluidized bed steam reforming (FBSR) as well as vitrification techniques based on cold crucible melting (CCM), Joule‐heater in‐container melting and plasma melting (PM) are described. Future developments in waste treatment will be based on separation technologies for partitioning individual RN along with design & construction of RN‐containing ceramic targets for inducing transmutation reactions. Near demonstration actinide‐hosting ceramic wasteforms including multiphase Synroc systems are described. Opportunities also exist for ceramics in structural applications in Generation IV reactors such as composite SiC/SiC and C/C for fuel cladding and control rods and MAX phases and ultrahigh‐temperature ceramics (UHTCs) may find near core fuel coating and cladding applications. Uses of ceramics in fusion reactor systems will be both functional (ceramic superconductors in magnet systems for plasma control and in Li silicate breeder blankets in tokamaks) and structural including as sapphire diagnostic windows, graphite diverters, and plasma facing C and UHTCs. In all these cases, performance is limited by poorly understood radiation damage and interface controlled processes, which demands a combined modeling/experimental approach.

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“…to provide feedback to fuel fabrication and modeling. 73,74 Mechanical interaction between layers has an impact on integrity of TRISO fuel particle, ultimately impacting its fission product retention characteristics. The ability to assess fission product retention and mechanical integrity of the fuel relies on the knowledge of temperature distribution which requires determining the conductivity for individual layers.…”

Section: Measurement In Compositesmentioning

confidence: 99%

“…72 Tests showed that TRISO fuel can maintain structural integrity, withstand irradiation and very high temperature up to 1000°C. 73 Extensive post irradiation examinations were done to characterize mechanical properties, chemical composition change, fission gas leakage, etc. to provide feedback to fuel fabrication and modeling.…”

Section: Measurement In Compositesmentioning

confidence: 99%

Investigation of thermal transport in composites and ion beam irradiated materials for nuclear energy applications

et al. 2016

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Thermal transport in materials used for energy applications is a physical process directly tied to performance and reliability. As a result, a great deal of effort has been devoted to understanding thermal transport in materials whose ability to conduct heat is critical. Here, our objective is to discuss the utility of laser-based thermoreflectance (TR) approaches that provide microscale measurement of thermal transport. We provide several examples that implement the TR technique to investigate thermal transport in materials used in nuclear energy applications. First, we discuss utility of this technique to measure thermal conductivity in ion irradiated ceramic materials during investigations where the primary objective is to understand the impact of radiation induced crystalline structure defects on thermal transport. We also present the capability of TR approach to resolve thermal conductivity of each layer in tristructural isotropic fuel, silicon carbide fiber composites, and 2nd phase precipitates in uranium silicide. Finally, the ability to measure interface thermal resistance between adjacent layers in composites is demonstrated.

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TRISO-Coated Particle Fuel Performance

Cited by 31 publications

References 25 publications

Safety evaluation of accident-tolerant FCM fueled core with SiC-coated zircalloy cladding for design-basis-accidents and beyond DBAs

Safety evaluation of accident-tolerant FCM fueled core with SiC-coated zircalloy cladding for design-basis-accidents and beyond DBAs

Opportunities for Advanced Ceramics and Composites in the Nuclear Sector

Investigation of thermal transport in composites and ion beam irradiated materials for nuclear energy applications

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