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Effect of Delays in Afterheat Removal on Consequences of Massive Air Ingress Accidents in High-Temperature Gas Cooled Reactors
Abstract
Massive ingress of air into the core of a high-temperature gas cooled reactor is among the accidents with a low occurrence frequency, but there are still gaps in understanding with respect to its consequences. In the present paper, massive air ingress combined with a delayed start of the afterheat removal system is investigated and compared to air ingress accidents with normal operation of the afterheat removal procedure. A computer programme REACT/THERMIX used for these accident analyses is described. For a high-temperature gas cooled reactor with a pebble bed core, it is shown that massive air ingress has no real safety endangering consequences even if the operation of the afterheat removal system is delayed by 6 h.
... In this context, SiC has garnered interest in extreme environments, including nuclear reactors, due to its high mechanical strength, thermal conductivity, and chemical stability [13]. With the proposed sleeveless fuel compacts design (Figure 2), the fuel matrix needs great anticorrosion properties to protect the TRISO fuel particles embedded in the fuel matrix, particularly in the event of an air ingress accident arising from a guillotine-type break of the main coolant pipe [14][15][16][17] that leads to severe mechanical and material degradation of the core structures and fuel compacts due to high temperature oxidation [18][19][20][21]. In the previous fuel design, the graphite sleeve was intended to serve as a chemical buffer against oxidation; however, in the sleeveless fuel design, the fuel matrix is naked and exposed to gases. ...
... In this context, SiC has garnered interest in extreme environments, including nuclear reactors, due to its high mechanical strength, thermal conductivity, and chemical stability [13]. With the proposed sleeveless fuel compacts design (Figure 2), the fuel matrix needs great anticorrosion properties to protect the TRISO fuel particles embedded in the fuel matrix, particularly in the event of an air ingress accident arising from a guillotine-type break of the main coolant pipe [14][15][16][17] that leads to severe mechanical and material degradation of the core structures and fuel compacts due to high temperature oxidation [18][19][20][21]. In the design of an accident-tolerant fuel (ATF), fully ceramic microencapsulated (FCM) fuel [22][23][24][25][26][27][28] utilizes SiC ceramics for their excellent ability to protect the fuel matrix during LWR accident conditions. ...
Preventing severe corrosion incidents caused by air ingress accidents in high-temperature gas-cooled reactors (HTGRs) while improving heat removal efficiency from the core is of paramount importance. To enhance both safety and efficiency, a sleeveless silicon carbide (SiC)-matrix fuel compact has been proposed. This study evaluates the 10-hour oxidation of reaction-sintered SiC (RS-SiC)-matrix fuel compact under the conditions of an air ingress accident within the temperature range of 1000 to 1400 °C. The oxidation tests were conducted in a stagnant air environment without flow. As a result, it is demonstrated that RS-SiC exhibits exceptional resistance to air oxidation up to 1400 °C, as shown by the thermogravimetric analysis (TGA), with minimal mass loss due to the oxidation of free carbon. Scanning electron microscopy with energy-dispersive X-Ray spectroscopy (SEM–EDX) analysis reveals that the morphology and thickness of the SiO2 layer formed on the RS-SiC surface vary with temperature. At 1400 °C, uniform oxide layer thickness ranging from 1.59 to 4.10 μm and localized nodule-like oxide formations of approximately 10 μm are observed. In contrast, at 1000–1200 °C, thinner oxide layers are identified, indicating that oxide growth accelerates at higher temperatures. The oxidation rates measured provide insights into the mechanisms of oxide growth.
... According to Eto and Growcock [4]'s research , the mechanical strength of nuclear graphite is significantly changed by the variation of graphite density, and as a result, approximately 10 percent density change may cause around 50 percent mechanical strength reduction depending on the materials. According to other graphite oxidation research (Fuller and Okho [5], Hino and Hishida [6], Hinsen et al. [7], Katcher and Moorman [8], Kim and NO [9], Moorman [10]), it takes around 7 or 8 days for all the graphite to be transformed into CO or CO 2 gas under 700 C oxidation condition, compared to 2 ~ 3 days over 900 o C. In reality, the graphite temperature can be increased up to 1500 o C in the accident conditions, and it means that the mechanical strength of the graphite can be seriously weakened causing the core collapse. ...
An air-ingress accident in a VHTR is anticipated to cause severe changes of graphite density and mechanical strength by oxidation process resulting in many side effects. However, quantitative estimations have not been performed yet. In this study, the focus has been on the prediction of graphite density change and mechanical strength using a thermal hydraulic system analysis code. For analysis of the graphite density change, a simple graphite burn-off model was developed based on the similarity concept between parallel electrical circuit and graphite oxidation considering the overall changes of the graphite geometry and density. The developed model was implemented in the VHTR system analysis code, GAMMA, along with other comprehensive graphite oxidation models. GT-MHR 600 MWt reactor was selected as a reference reactor. From the calculation, it was observed that the main oxidation process was derived 5.5 days after the accident following natural convection. The core maximum temperature reached up to 1400 o C. However it never exceeded the maximum temperature criteria, 1600 o C. According to the calculation results, most of the oxidation occurs in the bottom reflector, so the exothermic heat generated by oxidation did not affect the core heat up. However, the oxidation process highly decreased the density of the bottom reflector making it vulnerable to mechanical stress. In fact, since the bottom reflector sustains the reactor core, the stress is highly concentrated on this part. The calculations were made for up to 11 days after the accident and 4.5% of density decrease was estimated resulting in 25% mechanical strength reduction.
... CO and O2 homogeneous reaction (R4) happens in the boundary layer, which is the important secondary reaction in the oxidation processes of graphite. THERMIX/REACT [6]and TINTE [3,7,8] are currently widely used reactor physics and thermal hydraulic system codes for HTGR. In the oxidation models of these codes, the homogeneous reaction (R4) has been considered. ...
Gas chromatography and numerical methods were used to investigate the oxidation behavior of SNG742 nuclear grade graphite at 500–1100°C with different oxidizing gas flow rates. A four-step overall reaction mechanism was used to describe the graphite oxidation reaction process, which reasonably explains the peak of CO mole fraction at around 700°C with lower flow rates. The kinetic parameters of apparent oxidation reaction and four-step reaction mechanism of SNG742 graphite were obtained through curve fitting of the measured data. The apparent activation energy of SNG742 nuclear graphite is 200.7 ± 14.2 kJ/mol. The effects of temperature and gas velocity on the oxidation rate were analyzed. The graphite oxidation process was effectively simulated using CFD method with porous media model and the four-step reaction mechanism. The simulation results show that increasing the gas flow rate can reduce the thickness of the diffusion boundary layer and influence the diffusion control regime of graphite oxidation. The four-step overall reaction mechanism can better describe the graphite oxidation process than the one-step overall reaction. The porous medium model can effectively simulate the mass diffusion process and surface chemical reaction in the interior porous region of the graphite.
... Similar observations have been made previously looking at char oxidation with an interesting overview carried out by Shaddix et al [42]. In agreement with this present study, experiments carried out by Takahashi et al [43] on nuclear graphite demonstrated a rapid increase in CO/CO 2 at around 1000˚C, and postulated that the reason for the results was an increase in the endothermic Boudouard reaction (C +CO 2 ), which is favoured at higher temperatures [44]. Du et al [45] suggested that the general increase in CO/CO 2 with temperature is due to the higher activation energy for CO production compared to that of CO 2 . ...
This study has investigated the laboratory scale thermal oxidation of nuclear graphite, as a proof-of-concept for the treatment and decommissioning of reactor cores on a larger industrial scale. If showed to be effective, this technology could have promising international significance with a considerable impact on the nuclear waste management problem currently facing many countries worldwide. The use of thermal treatment of such graphite waste is seen as advantageous since it will decouple the need for an operational Geological Disposal Facility (GDF). Particulate samples of Magnox Reactor Pile Grade-A (PGA) graphite, were oxidised in both air and 60% O2, over the temperature range 400–1200°C. Oxidation rates were found to increase with temperature, with a particular rise between 700–800°C, suggesting a change in oxidation mechanism. A second increase in oxidation rate was observed between 1000–1200°C and was found to correspond to a large increase in the CO/CO2 ratio, as confirmed through gas analysis. Increasing the oxidant flow rate gave a linear increase in oxidation rate, up to a certain point, and maximum rates of 23.3 and 69.6 mg / min for air and 60% O2 respectively were achieved at a flow of 250 ml / min and temperature of 1000°C. These promising results show that large-scale thermal treatment could be a potential option for the decommissioning of graphite cores, although the design of the plant would need careful consideration in order to achieve optimum efficiency and throughput.
... Such increase in temperature increases the reaction rate constants, and the mobility and diffusion of the oxygen molecules and other oxidants from the bulk gas to the graphite surface. The heterogeneous corrosion reactions are: 23,34 ) ...
A massive air or steam ingress in High Temperature Reactors (HTRs) nominally operating at 600-950 o C is a design-basis accident requiring the development and validation of graphite oxidation and erosion models to examine the impact on the potential fission products release and the integrity of the graphite core and reflector blocks. Nuclear graphite is of many types with similarities but also differences in the microstructure, volume porosity, impurities, type and size of filler coke particles, graphitization and heat treatment temperatures, and the thermal and physical properties. These as well as the temperature, types and partial pressures of oxidants affects the prevailing oxidation mode and kinetics of the oxidation processes of graphite in HTRs. This paper reviews the graphite crystalline structure, the fabrication procedures, characteristics, chemical kinetics and modes of oxidation of nuclear graphite for future model developments.
... According to Eto and Growcock [4]'s research , the mechanical strength of nuclear graphite is significantly changed by the variation of graphite density, and as a result, approximately 10 percent density change may cause around 50 percent mechanical strength reduction depending on the materials. According to other graphite oxidation research (Fuller and Okho [5], Hino and Hishida [6], Hinsen et al. [7], Katcher and Moorman [8], Kim and NO [9], Moorman [10]), it takes around 7 or 8 days for all the graphite to be transformed into CO or CO 2 gas under 700 C oxidation condition, compared to 2 ~ 3 days over 900 o C. In reality, the graphite temperature can be increased up to 1500 o C in the accident conditions, and it means that the mechanical strength of the graphite can be seriously weakened causing the core collapse. ...
... Basic studies on the reaction mechanism of carbon and reactive gases have been carried out by Walker et al. [3]. While the reactions of nuclear grade graphite with oxygen and steam have been studied [4,5], these studies were done at temperatures below 1200 "C, and at steam concentrations of less than 5%. A more recent work used high concentration reactant gases at high temperatures [6]; the graphite-steam reaction rate was measured at temperatures between 1000 and 1700 "C, and the quantity of hydrogen generated during an accident involving coolant leakage into the plasma chamber was evaluated. ...
To obtain fundamental data for safety analyses of fusion reactors with regard to loss of coolant inside the vacuum vessel, the rate of corrosion reaction between high temperature graphite and steam was measured experimentally between 1000 and 1600 °C. A preliminary experiment gave an activation energy for reaction between oxygen and isotropic graphite as 45 kJ mol−1. The reaction rate with steam depended on temperature, with an inflection point near 1300 °C. The activation energy was about 270 kJ mol−1 at temperatures below 1300 °C, and 104 kJ mol−1 at higher temperatures. The energy was not strongly dependent on the graphite properties, but the reaction rate varied with them. The reactivity of the isotropic graphite was more than twice as high as that of C/C composite. The molar ratio of the product gases [H2]/[CO] increased slightly with increasing temperature. Although the [CO]/[CO2] ratio also increased with temperature to 1300 °C, it decreased above this temperature. This behavior reflects a change in reaction mechanism near this temperature; the composition of the product gas could be estimated numerically using elementary reaction rates.
Bei jedem Spaltvorgang im Kernreaktor entstehen neben den erwünschten Produkten Energie und Spaltneutronen Begleitstoffe, die den Anlagenbetrieb erschweren. Dies sind die Spaltprodukte, von denen mehrere hundert mit ihren Ausbeuten bei der Spaltung, ihren Halbwertzeiten, Zerfallsarten und Zerfallsenergien bekannt sind. Die Häufigkeit ihres Auftretens ist für die einzelnen Spaltstoffe charakteristisch und war bereits in Abb. 3.3, z. B. für U235 wiedergegeben. Bei jedem Spaltereignis werden zwei Spaltprodukte gebildet, die fast immer radioaktiv sind. Im allgemeinen entstehen Ketten, in denen konsekutive Zerfälle bis zum Erreichen eines stabilen Zustandes auftreten. Als einige Beispiele seien genannt: (6.1)
(6.2) Die Spaltprodukte [6.1 bis 6.3] sind von größter Wichtigkeit für den Reaktorbetrieb sowie für das Störfallgeschehen. In Abb. 6.1 sind einige wesentliche Gesichtspunkte, die auf die Bedeutung der Spaltprodukte für die verschiedenen Aspekte hinweisen, aufgelistet. In den folgenden Abschnitten soll insbesondere das Verhalten der Spaltprodukte im Hinblick auf Freisetzungen im Normalbetrieb und in Störfällen untersucht werden. Zur Bedeutung der Spaltprodukte im Hinblick auf die Zwischenlagerung oder die Endlagerung finden sich einige Ausführungen in Kap. 9.
A review of available correlations for graphite oxidation by steam, relevant in case of water ingress accidents in HTR, was performed. Accuracy of the available correlation for the primary reaction was investigated by performing comparison with available data. A new correlation was proposed. The new correlation is an extension of the existing correlation of Kubaschewski-Heinrich, obtained by introducing a pressure dependent and burn-off dependent multiplier for best estimate as well as conservative calculations. The best estimate correlation shows significant improvement of accuracy when compared with the same data. The maximum error and applicability range of the best estimate correlation are provided. The maximum deviation of the new correlation results and experimental data is -27% and +15%, compared to the maximum deviations of -72% and +76%, found for the original correlation. The range of application of the correlation is: 1000 K divided by 1300 K, 10(5) Pa divided by 55 x 10(5) Pa total pressure and 14 divided by 10(5) Pa steam partial pressure. The work presented in this paper was done within the ARCHER project.
We developed a multidimensional G As Multicomponent Mixture Analysis (GAMMA) code in order to investigate chemical reaction behaviors related to an air ingress accident and the thermofluid transients in high-temperature gas-cooled reactors. The implicit continuous Eulerian technique is adopted for the reduction of a 10N × 10N matrix into an N × N pressure difference matrix and fast transient computation. In the validation with a high-temperature engineering test reactor (HTTR)-simulated air ingress experiment, the onset times of natural convection are accurately predicted within a 10% deviation. Small internal leaks in the HTTR-simulated test facility have been found to significantly affect the consequence of air ingress. In all the simulated cases for a SANA-1 afterheat removal test, the predictions of GAMMA are in a high level of agreement with the measured temperature profiles and are comparable to the results of other codes (TINTE, THERMIX/DIREKT, and TRIO-EF).
Air ingress into to the core after the primary circuit depressurization due to large breaks of the pressure boundary is considered as one of the severe hypothetical accidents for the high temperature gas-cooled reactor (HTR). If the air source and the natural convection cannot be impeded, the continuous graphite oxidation reaction along with the formation of burnable gas mixtures resulting in the corrosion of the fuel elements and the reflectors might damage the reactor structure integrity and endanger the reactor safety. In order to study the effects of air flow driven by natural convection as well as to investigate the corrosion of graphite, the NACOK (Naturzug im Core mit Korrosion) facility was built at Jülich Research Center in Germany. A complete 2A-rupture of the coaxial duct in the HTR primary system, as well as the chimney effect caused by breaks in both upper and lower parts of the pressure boundary was simulated in the test facility. Several series of experiments and the related code validations (TINTE, DIREKT, THERMIX/REACT, etc.) have been performed on this facility since the 1990s. In this paper, the latest NACOK air ingress experiment, carried out on October 23, 2008 to simulate the chimney effect, was preliminarily analyzed at NRG with the SPECTRA code, as well as at INET, Tsinghua University of China with the TINTE code. The calculating results of air flow rate of natural convection, time-dependent graphite corrosion, and temperature distribution are compared with the NACOK test results. The preliminary code-to-experiment and code-to-code validation successfully proves the code capability to simulate and predict the air-ingress accident. In addition, more research work, including parameter sensitivity analysis, modeling refinement, code amelioration, etc., should be performed to improve the simulation accuracy in the future.
The oxidation behaviors of IG-110 and PGX graphites were experimentally investigated in the temperature range from 500 to 1,500°C for the oxygen concentration of 10 and 30 wt% (1.37 and 5.09 mol%) and the reacting gas velocities of 6.63 and 19.9 m/s (estimated at 900°C). The graphite specimens were prepared to simulate the annular coolant channel in the core of the HTGR. It has been found that the upstream surface is more oxidized for both IG-110 and PGX graphites, and that the oxidized surface of PGX is rougher than that of IG-110. The oxidation of IG-110 starts at about 700°C and reaches the maximum above 800 to 900°C accompanied by the maximum production of CO2 around 850°C and a monotonic increase in CO with increasing temperature. The oxidation behavior of PGX is similar to that for IG-110, except that the oxidation starts at lower temperature and produces more CO2 over 750 to 950°C while the concentration of CO becomes lower in this temperature range. The present result for the product ratio of CO to CO2 deviates from the linear line in the Arrhenius plot in the region where the production of CO2 increases. However, it agrees best with the Rossberg correlation among the previous empirical correlations. New empirical correlations for the product ratio of CO to CO2 have been obtained from the present data which exclude those in the region of CO2 increase.
This study focuses on predicting the changes in graphite density and mechanical strength in VHTR during the air-ingress accident via thermal hydraulic system analysis code. A simple graphite burn-off model was developed based on the similarities between a parallel electrical circuit and graphite oxidation. The developed model along with other comprehensive graphite oxidation models were integrated into the VHTR system analysis code, GAMMA. GT-MHR 600MWt reactor was selected as a reference reactor. Based on the calculation, the main oxidation process was observed 5.5 days after the accident when followed by natural convection. The core maximum temperature reached 1430°C, but never exceeded the maximum temperature criteria, 1600°C. However, the oxidation process did significantly decrease the density of bottom reflector, making it vulnerable to mechanical stress. The stress on the bottom reflector is greatly increased because of the reduction of loaded surface area with graphite oxidation. The calculation proceeded until 11 days after the accident, resulting in an observed 4.5% decrease in density and a 25% reduction of mechanical strength.
An experimental study on mass transfer with chemical reactions and in-pore diffusion was performed in cross flow containing oxygen past a porous graphite cylinder placed in a high temperature environment. Reynolds numbers ranged from 533 to 2490, and cylinder temperatures ranged from 848 to 1120°C. It is concluded that chemical reactions do not significantly influence the mass transfer, and that the effect of in-pore diffusion on mass transfer can be estimated by using empirical relations for the corrosion rate and mass transfer.
The advanced modular high temperature gas-cooled reactor, currently being developed in the United States, employs several inherent and passive safety systems. To assess its safety potential under very severe and low probability accident scenarios, the analysis tools for such transients have been developed and are applied. Applications of the resulting codes to these accident scenarios are presented, showing that even these severe accident transients would not cause any significant fission product releases, but might lead to investment losses.
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