Ingress Accident Research Articles

Correlation for steam–graphite reaction

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: 1000K÷1300K, 105Pa÷55×105Pa total pressure and 14÷105Pa steam partial pressure. The work presented in this paper was done within the ARCHER project.

Oxidation rate of graphitic matrix material in the kinetic regime for VHTR air ingress accident scenarios

Data on oxidation rates of matrix-grade graphite in the kinetically-controlled temperature regime of graphite oxidation are needed for safety analysis of High Temperature Gas Cooled Reactors and Very High Temperature Reactors. In this work, the oxidation rate of graphitic matrix material GKrS was measured thermogravimetrically for various oxygen concentrations and with temperatures from 873 to 1873K. A semi-empirical Arrhenius rate equation was also developed for this temperature range. The activation energy of the graphitic material is found to be about 111.5kJ/mol. The order of reaction was found to be about 0.89. The surface of oxidized GKrS was characterized by Scanning Electron Microscopy, Electron Dispersive Spectroscopy, Fourier Transform Infrared Spectroscopy and X-ray Photoelectron Spectroscopy.

Feasibility study on naturally safe HTGR (NSHTR) for air ingress accident

A new design activity for an advanced reactor, referred to as a naturally safe high temperature gas-cooled reactor (NSHTR), has been launched by authors after the accident at the Fukushima Daiichi Nuclear Power Station. The concept of NSHTR is that the release of radioactive materials is kept at very low level and no harmful effect on people and the environment is ensured by only physical phenomena even in the absence of engineered safety features. At an air ingress accident, possible physical events that lead to the loss or degradation of the confinement function of the fuel-coating layers are the crack of the coatings caused by the explosion of carbon monoxide (CO) produced by the graphite oxidation and failure of the coatings by melting or sublimation caused by core heat up due to the reaction heat of the graphite oxidation. In this study, the CO concentration and the heat generated by graphite oxidation inside the circular tube were evaluated parametrically using a steady-state one-dimensional model to confirm the feasibility of NSHTR at a severe condition of the air ingress accident (i.e., a massive air ingress by simultaneous rupture of two primary pipes). It was confirmed that the CO concentration at the outlet of coolant channel can be maintained below the explosion limit due to the reaction with oxygen in the air, and the reaction heat can be removed with the decay heat by physical phenomena under certain conditions of the coolant channel geometry without any engineered safety features. The results revealed that the design of NSHTR is technically feasible in terms of the suppression of the CO explosion and the heat removal of the reaction heat at the air ingress accident.

Analysis of diffusion process and influence factors in the air ingress accident of the HTR-PM

Air ingress, one of the beyond design basis accidents for high temperature gas-cooled reactors, receives high attention during the design of the 250MW pebble-bed modular high temperature gas-cooled reactor (HTR-PM), because it may result in severe consequence including the corrosion of the fuel element and graphite reflector. The diffusion process and the set-up time of the stable natural convection after the double-ended guillotine break of the hot-gas duct are studied in the paper. On the basis of the preliminary design of the HTR-PM and its DLOCA analysis results, the diffusion process, as well as the influence of the core temperature distribution and the length of the hot-gas duct, is studied with the DIFFLOW code, which adopts a one-dimension variable cross-section diffusion model with fixed wall temperature. To preliminarily estimate the influence of chemical reaction between oxygen and graphite, which will change the gas component of the mixture, the diffusion processes between the He/N2, He/O2, He/CO and He/CO2 are calculated, respectively. Furthermore, the code has been improved and the varying wall temperature can be simulated. The more accurate analysis is carried out with the changing temperature distribution from the DLOCA calculation. The analysis shows that there is enough time to adopt appropriate mitigation measures to stop the air ingress and the severe consequence of fuel element damage and large release of fission product can be avoided.

Analysis on blow-down transient in water ingress accident of high temperature gas-cooled reactor

Water ingress into the primary circuit is generally recognized as one of the severe accidents with potential hazard to the modular high temperature gas-cooled reactor, which will cause a positive reactivity introduction with the increase of steam density in reactor core to enhance neutron slowing-down, also the chemical corrosion of graphite fuel elements and the damage of reflector structure material. The increase of the primary pressure may result in the opening of the safety valves, consequently leading the release of radioactive isotopes and flammable water gas. The research on water ingress transient is significant for the verification of inherent safety characteristics of high temperature gas-cooled reactor. The 200MWe high temperature gas-cooled reactor (HTR-PM), designed by the Institute of Nuclear and New Energy Technology of Tsinghua University, is exampled to be analyzed in this paper. The design basis accident (DBA) scenarios of double-ended guillotine break of single heat-exchange tube (steam generator heat-exchange tube rupture) are simulated by the thermal-hydraulic analysis code, and some key concerns which are relative to the amount of water into the reactor core during the blow-down transient are analyzed in detail. The results show that both of water mass and steam ratio of the fluid spouting from the broken heat-exchange tube are affected by break location, which will increase obviously with the broken location closing to the outlet of the heat-exchange tube. The double-ended guillotine rupture at the outlet of the heat-exchange will result more steam penetrates into the reactor core in the design basis accident of water ingress. The mass of water ingress will also be affected by the draining system. It is concluded that, with reasonable optimization on design to balance safety and economy, the total mass of water ingress into the primary circuit of reactor could be limited effectively to meet the safety requirements, and the pollution of radioactive isotopes releasing to the secondary circuit during the blow-down transient will be prevented.

R&D plan for development of oxidation-resistant graphite and investigation of oxidation behavior of SiC coated fuel particle to enhance safety of HTGR

A new concept of the high temperature gas-cooled reactor (HTGR) is proposed as a challenge to assure no event sequences to the harmful release of radioactive materials even when the design extension conditions (DECs) occur by deterministic approach based on the inherent safety features of the HTGR. One of the DECs for the HTGR is the air/water ingress accidents. As the present status of the HTGR, such as the High Temperature Engineering Test Reactor (HTTR) in Japan Atomic Energy Agency (JAEA), graphite sleeves hold the fuel compact which consists of the SiC-TRISO (TRIstructural-ISOtropic) coated fuel particle, and make coolant flow channels. Then, the air/water ingress accident at the HTGR is prevented by additional measures (e.g. facility for suppression to air ingress). With regard to the core design, it is important to prevent recriticality accidents by keeping the geometry of the fuel rod which consists of the graphite sleeve, fuel compact and SiC-TRISO coated fuel particle, and by improving the oxidation resistance of the graphite when air/water ingress accidents occur. Therefore, it is planned to develop the oxidation-resistant graphite, which is coated with gradient SiC layer. It is also planned that the experimental identification of the condition to form the stable oxide layer (SiO2) for SiC layer on the oxidation-resistant graphite and on the SiC-TRISO fuel. This paper describes the R&D plan for un-irradiation and irradiation test under simulating air/water ingress accident condition to develop oxidation-resistant graphite and to investigate the oxidation behavior of SiC coated fuel particle.

Development of evaluation method with X-ray tomography for material property of IG-430 graphite for VHTR/HTGR

Graphite materials are used for the in-core components of High Temperature Gas-cooled Reactor (HTGR) which is a graphite-moderated and helium gas-cooled reactor. The HTGR is particularly attractive due to capability of producing high temperature helium gas, and its passive and inherent safety features. The Very High Temperature Reactor (VHTR) is one of the most promising candidates for the Generation-IV nuclear reactor systems. IG-110 graphite having high strength and resistance to oxidation is used in the HTTR of JAEA. IG-110 is a major candidate for the in-core graphite components of VHTR, too. From the standpoint of the safety at air ingress accident, it is important for graphite materials to have adequate resistance against oxidation damage. IG-430 graphite having higher strength and resistance to oxidation than IG-110 is an advanced candidate for the VHTR. Recently, X-ray tomography method is expected to apply the evaluation of neutron irradiation effects by measuring the irradiation-induced change of geometry of graphite grains and pores. This method is also applicable to evaluate the oxidation damage on graphite from the oxidation-induced change of grain/pore microstructures. In this study, in order to develop evaluation method for material properties and to evaluate the irradiation-induced property changes under higher neutron doses for IG-430, the oxidation and densification effects on elastic modulus of IG-430 were investigated. Moreover, the correlation of the microstructure based on the X-ray tomography images and the material properties was discussed. It was shown that the elastic modulus of the densified graphite depends on only the open pores and it is possible to evaluate the material properties of graphite by using X-ray tomography method. However, it is necessary to take into account of the change in the number and shape of closed pores in the grain to simulate the elastic modulus of the highly oxidized and irradiated materials by the homogenization analysis.

Oxidation rate of nuclear-grade graphite IG-110 in the kinetic regime for VHTR air ingress accident scenarios

The oxidation rates of nuclear-grade graphite IG-110 in the kinetically-controlled temperature regime of graphite oxidation were predicted and compared in Very High Temperature Reactor air ingress accident scenarios. The oxidative mass loss of graphite was measured thermogravimetrically from 873 to 1873K in 100% air (21mol%). The activation energy was found to be 222.07kJ/mol, and the order of reaction with respect to oxygen concentration is 0.76. The surfaces of the samples were characterized by Scanning Electron Microscopy, Energy Dispersive Spectroscopy, Fourier Transform Infrared Spectroscopy and X-ray Photoelectron Spectroscopy before and after oxidation. These results are compared with those available in the literature, and our recently reported results for NBG-18 nuclear-grade graphite using the same technique.

Oxidation rate of nuclear-grade graphite NBG-18 in the kinetic regime for VHTR air ingress accident scenarios

One of the most severe accident scenarios anticipated for VHTRs is an air ingress accident caused by a pipe break. Graphite oxidation could be severe under these conditions. In this work, the oxidation rate of NBG-18 nuclear-grade graphite was studied thermogravimetrically for different oxygen concentrations and with temperatures from 873 to 1873K. A semi-empirical Arrhenius rate equation was developed for the temperature range of 873–1023K. The activation energy of NBG-18 was 187kJ/mol and the order of reaction was 1.25. The penetration depth of oxidant was about 3–4mm for NBG-18 oxidized at 973K. Increased porosity and changes in external geometry became more prominent at higher temperatures from about 1173 to 1873K. The surface of oxidized NBG-18 was characterized by SEM, EDS, FTIR and XPS.

Chemical kinetics parameters of nuclear graphite gasification

This paper provides chemical kinetics parameters for the gasification of nuclear graphite grades of IG-110, IG-430, NBG-18 and NBG-25 and presents empirical correlations for their surface areas of free active sites as a function of mass. The kinetics parameters for the four elementary chemical reactions of gasification of these grades of nuclear graphite include the values and Gaussian distributions of the specific activation energies and the values of the pre-exponential rate coefficients for the adsorption of oxygen and desorption of CO and CO2 gases. The values of these parameters and the surface area of free active sites for IG-110 and NB-25, with fine and medium petroleum coke filler particles, are nearly the same, but slightly different from those of NBG-18 and IG-430, with medium and fine coal tar pitch coke filler particles. Recommended parameters are applicable to future safety analysis of high and very high temperature gas cooled reactors in the unlikely event of a massive air ingress accident.

Effects of the onset time of natural circulation on safety in an air ingress accident involving a HTGR

An air ingress accident is a major safety issue pertaining to high-temperature gas-cooled reactors (HTGRs). When a helium depressurization accident occurs, air enters the core through broken pipes. The natural circulation (NC) phase of this accident leads to massive air ingress, causing extensive oxidation of the core graphite structure. The oxidation results in three main safety issues: an increase in the core temperature, degradation of the core graphite structure, and explosive CO accumulation in the upper plenum. As a result, the onset time of the natural circulation is one of the key parameters related to reactor safety. Using GAMMA (GAs Multi-component Mixture transient Analysis) we performed a one-dimensional (D) analysis of GT-MHR in the same manner used by the conventional safety tools. A 2D analysis was also done to determine the 2D effects, including the degree of stratified-flow induced-air ingress in the horizontal ducts. The 1D and 2D simulations showed that NC occurs at around 360h and over 500h, respectively (Jin et al., 2011). In an effort to determine the effect of the onset time of NC on safety, we performed a sensitivity study of the NC onset time when it is intentionally controlled. Due to the limitation of the 1D simulation to represent multi-dimensional stratified flows in hot and cold pipes, the 1D and 2D results are quite different. In the 1D calculation, the bottom reflector temperature is very sensitive to the onset time of NC, whereas it is insensitive in the 2D simulation. According to the 2D analysis, however, a strong-back flow from the air in the cavity is observed in the lower part of the reactor core and at the entrance of the broken pipes. As a result, the bottom reflector temperature gradually decreases in the 2D analysis irrespective of the onset time of NC. Based on these calculation results, the possibility of CO detonation and damage to the graphite structure are examined. Our conclusions are that there are large safety margins of the maximum fuel temperature, CO detonation, and damage to the graphite structure and even larger margins considering the more realistic 2D simulation.

Conceptual study on air ingress mitigation for VHTRs

An air ingress accident following a postulated pipe break is considered a critical event for a very high temperature gas-cooled reactor (VHTR) safety. Following helium depressurization, it is anticipated that air will enter the core through the break leading to oxidation of the in-core graphite structures. Under extreme circumstances and without mitigation features this accident may lead to exothermic chemical reactions between graphite and oxygen depending on the accident scenario and the design. Under extreme circumstances (beyond design basis), a loss of structural integrity may occur in some core structures and lead to elevated release of radiological inventory for the fuel matrix.This paper discusses various air ingress mitigation concepts applicable for the VHTRs that would prevent core damage even in the most extreme scenarios. The study begins with identifying important factors (or phenomena) associated with the air ingress accident using root-cause analysis. By preventing main causes of the important events identified in the root-cause diagram, the basic air ingress mitigation ideas were conceived and developed.Among them, two concepts were finally evaluated as effective candidates. One concept is to inject helium directly into the lower plenum (direct in-vessel injection); the other concept is to enclose the reactor with a non-pressure boundary with an opening at the bottom (ex-vessel enclosure). Computational fluid dynamics (CFD) methods were used to evaluate these concepts for proof of these principles. Results indicate that both concepts can effectively suppress air ingress. In the first concept, the buoyancy of the injected helium replaces the air in the core and the upper part of lower plenum. This prevents air from moving into the reactor core and shows the most potential for mitigating graphite oxidation within the vessel. In the second concept, air ingress is limited by molecular diffusion through the opening at the enclosure bottom, a very slow process that allows sufficient time for the core to cool. Existing VHTR concepts can be modified to accommodate either of these mitigation concepts.

CFD investigating the air ingress accident for a HTGR simulation of graphite corrosion oxidation

Abstract Through a compressible multi-component CFD model, this paper investigates the characteristics of graphite oxidation corrosion in the HTR-10 core under the postulated accident of gas duct rupture. In this accident, air in the steam generator cavity would enter into the core after pressure equilibrium is achieved between the core and the cavity, which is also called as the air ingress accident. Oxygen in the air would react with graphite on pebble surface, subsequently resulting in oxidation corrosion and challenging fuel integrity. In this paper, characteristics of graphite oxidation corrosion during the air ingress accident can be reasonably captured, including distributions of graphite corrosion amount on the different cross-sections, time histories of local corrosion amount at the monitoring points and overall corrosion amount in the core, respectively. Based on the transient simulation results, the corrosion pattern and its corrosion rate would approach to the steady-state conditions as the accident continuously progresses. The total amount of graphite corrosion during a 3-day accident time is predicted to be about 31 kg with the predicted asymptotic corrosion rate. This predicted value is less than that from the previous work of Gao and Shi ( Gao and Shi, 2002 ).

CFD investigating the air ingress accident occurred in a HTGR_simulation of thermal-hydraulic characteristics

Abstract The majority of this paper investigates thermal-hydraulic characteristics in a high-temperature gas-cooled reactor (HTGR) core under steady-state and accident conditions through a transient three-dimensional compressible CFD model. With the pebble-bed geometry, the HTR-10 core is selected for the present simulations. A loss-of-coolant accident (LOCA) with double-ended rupture of a gas duct tube is assumed and the core power would decay following this accident. In this LOCA, air in the steam generator cavity would enter the core via diffusion and natural convection mechanisms. Following air ingress into the core, O2 would react with graphite on pebble surface, causing the generation of CO2 and CO. The main thermal-hydraulic characteristics during this air ingress accident include O2 ingress into the core, CO2 and CO production histories, He concentration variations, and coolant temperature transient behavior, etc. Based on the present simulation results, the peak coolant temperature during the air ingress accident is slightly higher than that under the steady-state condition, revealing the safety of HTR-10 core in this postulated accident from the thermal point of view. In addition, corrosion phenomena related to the fuel pebbles are discussed and presented in the next paper.

Computational prediction of dust production in pebble bed reactors

Abstract This paper describes the computational modeling and simulation of graphite pebbles in frictional contacts as anticipated in a pebble bed reactor. For the high temperature gas-cooled reactor, the potential dust generation from frictional contact at the surface of pebbles and the subsequent lift-off and transport of dust and absorbed fission products are of safety concern at elevated temperatures under an air ingress accident. The aim of this work is to perform a computational study to estimate the quantity of the nuclear grade graphite dust produces from a typical anticipated configuration.

Effects of different surfaces on the transport and deposition of ruthenium oxides in high temperature air

In order to understand the behaviour of ruthenium oxides in the reactor coolant system during an air ingress accident, new tests were performed in the frame of the RUSET (RUthenium Separate Effect Test) experimental program. These aimed to ascertain the effects of different surfaces (quartz, stainless steel (SS), zirconium alloy, alumina, oxidised metal, and surfaces with Mo or Cs deposits) on the transport and decomposition of ruthenium oxides in air stream along the temperature gradient zone (1100–100 °C). The results demonstrated that the heterogeneous phase decomposition of RuO 3 and RuO 4 to RuO 2 is catalysed more efficiently by the quartz surface than by the SS or alumina surfaces. The presence of MoO 3 layers decreased the RuO x precipitation extent on all investigated surfaces. The trapping effect of Cs deposit on Ru in the temperature gradient zone was proved in the case of the SS surface. On the contrary, presence of Cs precipitate on alumina and especially on quartz surfaces was found to decrease their catalytic effect on the decomposition of ruthenium oxides, and thus increased the RuO 4 concentration in the outlet air. Similarly to the effect observed for Cs deposition, the presence of other fission products in the evaporation area (at 1100 °C) decreased the partial pressure of RuO 4 in the outlet air at the SS surface and increased it at quartz and alumina surfaces. When zirconium (E110) cladding material was placed in the temperature gradient zone, no Ru transmittance occurred until the high temperature end of the zirconium tube was completely oxidised. After the intense oxidation of E110, Ru release occurred only in the presence of other fission product species. Pre-oxidation of SS surfaces in steam had no significant effect on the Ru passage.

HTR-TN Achievements and Prospects for Future Developments

It is already 10 years since the (European) High Temperature Reactor Technology Network (HTR-TN) launched a program for development of HTR technology, which expanded through three successive Euratom framework programs, with many projects in line with the network strategy. Widely relying in the beginning on the legacy of the former European HTR developments (DRAGON, AVR, THTR, etc.) that it contributed to safeguard, this program led to advances in HTR/VHTR technologies and produced significant results, which can contribute to the international cooperation through Euratom involvement in the Generation IV International Forum (GIF). the main achievements of the European program, performed in complement to efforts made in several European countries and other GIF partners, are presented: they concern the validation of computer codes (reactor physics, as well as system transient analysis from normal operation to air ingress accident and fuel performance in normal and accident conditions), materials (metallic materials for vessel, direct cycle turbines and intermediate heat exchanger, graphite, etc.), component development, fuel manufacturing and irradiation behavior, and specific HTR waste management (fuel and graphite). Key experiments have been performed or are still ongoing, like irradiation of graphite and of fuel material (PYCASSO experiment), high burn-up fuel PIE, safety test and isotopic analysis, IHX mock-up thermohydraulic test in helium atmosphere, air ingress experiment for a block type core, etc. Now HTR-TN partners consider that it is time for Europe to go a step forward toward industrial demonstration. In line with the orientations of the “Strategic Energy Technology Plan (SET-Plan)” recently issued by the European Commission that promotes a strategy for development of low-carbon energy technologies and mentions Generation IV nuclear systems as part of key technologies, HTR-TN proposes to launch a program for extending the contribution of nuclear energy to industrial process heat applications addressing (1) the development of a flexible HTR that can be coupled to many different process heat and cogeneration applications with very versatile requirements, (2) the development of coupling technologies for such coupling, (3) the possible adaptations of process heat applications required for nuclear coupling, and (4) the integration and optimization of the whole coupled system. As a preliminary step for this ambitious program, HTR-TN endeavors to create a strategic partnership between nuclear industry and R&D and process heat user industries.

Phenomenology of Graphite Burning in Air Ingress Accidents of HTRs

Air ingress with graphite burning belongs to the accident scenarios in HTRs with potentially severe consequences. This paper gives an overview of basic phenomena of graphite burning like ignition conditions and moving reaction fronts. The pioneering graphite burning experiments of Don Schweitzer are successfully reevaluated. Ignition conditions are examined, and it is underlined that burning depends not only on graphite properties but also on the heat balance in the whole graphite arrangement. In graphite‐moderated reactors, ignition occurs at about 650°C for small air flow rates: this means that normal operation temperatures in HTRs always allow for ignition. Fuel behaviour in air ingress, as determined in the KORA facility, is discussed: up to about 1300°C modern TRISO fuel is stable in air, but from 1500°C a complete, fast destruction is observed. Exemplary calculations on massive air ingress by chimney draught performed with REACT/THERMIX are outlined. For a hot bottom reflector there is a substantial time span before fuel is attacked. Because severe air ingress in well‐designed HTRs belongs to beyond design basis accidents, the knowledge is fairly good. Concerning protecting measures, a more detailed examination of thick SiC layers is proposed.

Wastage and Self-Plugging by a Potential CO2 Ingress in a Supercritical CO2 Power Conversion System of an SFR

For a CO2 ingress accident into liquid sodium in a supercritical CO2 power conversion system coupled with a sodium-cooled fast reactor, we investigated two major design issues: i) a wastage phenomenon in regard to structural damage adjacent to the leaking position, and ii) potential channel plugging due to the formation of a particulate reaction product. In order to understand the factors affecting the occurrence of these issues, two kinds of experiments were carried out: a wastage effect test and a self-plugging test. All experimental conditions were chosen to reasonably represent the normal operating conditions and realistic design parameters of the reference plant. The test results indicate the absence of wastage, which will not lead to additional tube ruptures and damage propagation. In the current experiment, the self-plugging of PCHE channels only took place under two limited conditions: i) the sodium temperature is over 500°C and ii) the equivalent diameter of the crack opening is less than 1.5mm with a small leakage rate of far less than 1 g/s of CO2 ingress.

Wastage and Self-Plugging by a Potential CO2 Ingress in a Supercritical CO2 Power Conversion System of an SFR

For a CO2 ingress accident into liquid sodium in a supercritical CO2 power conversion system coupled with a sodium-cooled fast reactor, we investigated two major design issues: i) a wastage phenomenon in regard to structural damage adjacent to the leaking position, and ii) potential channel plugging due to the formation of a particulate reaction product. In order to understand the factors affecting the occurrence of these issues, two kinds of experiments were carried out: a wastage effect test and a self-plugging test. All experimental conditions were chosen to reasonably represent the normal operating conditions and realistic design parameters of the reference plant. The test results indicate the absence of wastage, which will not lead to additional tube ruptures and damage propagation. In the current experiment, the self-plugging of PCHE channels only took place under two limited conditions: i) the sodium temperature is over 500°C and ii) the equivalent diameter of the crack opening is less than 1.5mm w...