Oxidation of PCEA nuclear graphite by low water concentrations in helium

Contescu, Cristian I.; Mee, Robert W.; Wang, Peng; Romanova, A. V.; Burchell, Timothy D

doi:10.1016/j.jnucmat.2014.07.009

Cited by 31 publications

(39 citation statements)

References 22 publications

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“…This report summarizes the achievements and the current status of this multi-year research effort. It makes reference to already published studies [11,12,13,14] and contains updates on the latest results. The main goal is to emphasize the experimental progress so far and the challenges encountered with data analysis.…”

Section: Introductionmentioning

confidence: 99%

Status of Chronic Oxidation Studies of Graphite

Contescu

Mee

2016

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Graphite will undergo extremely slow, but continuous, oxidation by traces of moisture that will be present, albeit at very low levels, in the helium coolant of an HTGR. This chronic oxidation may cause degradation of mechanical strength and of other properties of graphite components if a porous oxidation layer penetrates deep enough in the bulk of graphite components during the lifetime of the reactor. The current research on graphite chronic oxidation is motivated by the acute need to understand the behavior of each graphite grade during prolonged exposure to chemical attack by moisture at high temperature. The goal is to provide the elements needed to develop predictive models for long-time oxidation behavior of graphite components in the cooling helium of HTGR. The tasks derived from this goal are: (1) Oxidation rate measurements in order to determine and validate a comprehensive kinetic model suitable for prediction of intrinsic oxidation rates as a function of temperature and oxidant gas composition; (2) Characterization of effective diffusivity of water vapor in the graphite pore system in order to account for the in-pore transport of moisture; and (3) Development and validation of a predictive model for the penetration depth of the oxidized layer, in order to assess the risk of oxidation-caused damage of particular graphite grades after prolonged exposure to the environment of helium coolant in an HTGR. The most important-and most time consuming-of these tasks is the measurement of oxidation rates in accelerated oxidation tests (but still under kinetic control) and the development of a reliable kinetic model. This report summarizes the status of chronic oxidation studies on graphite, and then focuses on model development activities, progress of kinetic measurements, validation of results, and improvement of the kinetic models. Analysis of current and past results obtained with three nuclear graphite grades showed that the classical Langmuir-Hinshelwood model cannot reproduce all data collected so far. Starting from here we propose a modification of the LH model to include temperature activation of the graphite surface, modeled as a Boltzmann activation function. The Boltzmann-enhanced Langmuir-Hinshelwood model (BLH) was tested successfully on three grades of graphite. The model is a robust, comprehensive mathematical function that allows better fitting of experimental results spanning a wide range of temperatures and partial pressures of water vapor and hydrogen. However, the model did not improve much the fitting of old data on graphite H-451 oxidation by water.

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Section: Introductionmentioning

confidence: 99%

Status of Chronic Oxidation Studies of Graphite

Contescu

Mee

2016

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“…Since the outermost layer of the particles consisted of the CVD-SiC coating layer, the particles were only adjacent to each other and the SiC tray; hence no possibility from direct interaction with foreign materials was possible during the tests. Particles without OPyC were intentionally used since it is known that graphite reacts rapidly in presence of water vapor [11] and also to isolate the study to that of SiC coating layer oxidation.…”

Section: Methodsmentioning

confidence: 99%

“…The majority of test conditions, consisting of 100% H 2 O at elevated temperatures (1500-1700°C), are unique and depart from what is considered applicable for HTGRs under normal and most accident conditions. Under normal operating conditions for HTGRs, only a small partial pressure of water vapor is expected to be present (<1.4 Pa [11]) and under severe accidents the combination of high water vapor partial pressure (10 5 Pa and higher) and elevated temperatures of this study are unlikely [12]. However, for the LWR application of TRISO fuel particles as microencapsulated fuel forms [13], namely fully ceramic microencapsulated (FCM) pellet, severe accidents beyond the design basis can lead to these high temperatures in presence of abundant steam.…”

Section: Introductionmentioning

confidence: 99%

High temperature steam oxidation of SiC coating layer of TRISO fuel particles

Terrani

Silva

2015

Journal of Nuclear Materials

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a b s t r a c tHigh-temperature oxidation behavior of SiC coating layer of TRISO fuel particles in 1500-1700°C steam at 1 atm has been examined inside a zirconia furnace. The SiC coating layers experienced a thickness loss of less than 2.5 lm under these conditions up to 24 h. The thickness of the oxide layer formed under these conditions was consistent with prior steam oxidation tests on high-purity bulk SiC. Upon reducing the presence volatile impurities from the test environment (particularly Al) by conducting the tests inside a zirconia furnace, melting of the silica layer at 1700°C was avoided.

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“…During operation, the concentration of contaminants in the coolant‐helium is maintained at a low level by cycling a certain fraction of the coolant through a purification system. However, the fuel elements may be oxidized by low concentration oxidants (eg, H 2 O, CO 2 , O 2 ) presented in the helium . The ingress of steam into the coolant from a damaged heat exchanger is probably the most likely accident which would destroy the integrity of the layers of TRISO particles .…”

Section: Introductionmentioning

confidence: 99%

“…However, the fuel elements may be oxidized by low concentration oxidants (eg, H 2 O, CO 2 , O 2 ) presented in the helium. 3 The ingress of steam into the coolant from a damaged heat exchanger is probably the most likely accident which would destroy the integrity of the layers of TRISO particles. 4 Once the graphite matrix of the compacts or pebbles has been oxidized away and the SiC shell of the TRISO is exposed, the steam oxidation would cause degradation of the mechanical and thermal properties of the outer layers and increase the failure probability, and in the worst case results in a rapid release of volatile fission products (eg, Xenon and Krypton).…”

Section: Introductionmentioning

confidence: 99%

Effects of water vapor on the oxidation and the fracture strength of SiC layer in TRISO fuel particles

Cao

Hao

Wang³

et al. 2017

J Am Ceram Soc

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Steam oxidation of silicon carbide (SiC) layer in nuclear fuel particles were performed in flowing argon‐water vapor mixture with a total pressure of 1 bar at 1173‐1673 K. Both the phase composition and the microstructure of the oxide scale on the SiC layer varied with the oxidation temperature. Reaction rates of water vapor with the SiC layer were determined by measuring the oxide scale thickness. It was found that the oxidation of SiC layer follows the parabolic law. The activation energy was calculated to be 103±11 kJ/mol. It is proposed that the rate determine step of the oxidation is the diffusion of water vapor molecules in the oxide scale. The fracture strength of SiC shell after steam oxidation was evaluated using a crush test. The fracture strength decreased with the increase in the oxidation temperature due to the thinning of the SiC layer.

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Oxidation of PCEA nuclear graphite by low water concentrations in helium

Cited by 31 publications

References 22 publications

Status of Chronic Oxidation Studies of Graphite

Status of Chronic Oxidation Studies of Graphite

High temperature steam oxidation of SiC coating layer of TRISO fuel particles

Effects of water vapor on the oxidation and the fracture strength of SiC layer in TRISO fuel particles

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