Air Ingress Scenarios Research Articles

Analysis of air ingress scenario in GEMINI+ plant

The work presented in this paper was performed within the Euratom Horizon 2020 GEMINI+ project. The GEMINI+ reactor is a prismatic block-type High-Temperature Gas-Cooled Reactor (HTGR). Within tasks devoted to safety, an analysis of air ingress scenario was performed by NRG with the thermal–hydraulic system code SPECTRA, while UJV and NCBJ worked with the integral system code MELCOR. Two air ingress scenarios were analyzed:Design Basis Accident (DBA) scenario. The air ingress scenario selected as DBA is a 65 mm break of a Helium purification line on top of the Steam Generator,Large air ingress scenario. The air ingress scenario selected is guillotine break of the coaxial gas duct.The DBA scenario is characterized by a very long period with practically stagnant gas in the primary system after the initial depressurization. However, very small gas flows through the break due to counter flow and diffusion in the break region (gas velocities of 10–2 m/s) are difficult to model with system codes. Therefore the break models in the system code were calibrated by performing CFD simulations. Furthermore, a model of gas mixing by diffusion was introduced in the NRG analysis for the practically stationary gas in the primary system (gas velocities of 10–4 m/s). The large air ingress scenario, with relatively large flow through the break, is easier to model from this standpoint.The NRG SPECTRA results showed that the amount of air that can reach the core is extremely small in the DBA accident (0.5 kg of graphite consumed after 100 h). The maximum depth of oxidation was 2 µm. In the large air ingress scenario the air ingress is significantly larger. At 100 h, approximately 105 kg of graphite was consumed. Even then, the maximum depth of oxidation was very small, approximately 0.2 mm. MELCOR results, performed at NCBJ and UJV, basically confirmed that the amount of air ingress is very small. The numbers were somewhat higher, mainly due to heavy flow oscillations in the break that could not be mitigated in MELCOR calculations.

Effect of steam and oxygen starvation on severe accident progression with air ingress

The CODEX-AIT-3 experiment simulated an in-vessel air ingress scenario after failure of bottom head of a nuclear reactor. 1600 °C maximum temperature was reached with the electrically heated bundle and slow cool-down was applied at the end of the test to demonstrate the bundle state before quench. Limited air and steam flow rates were intentionally selected in order to slow down the high temperature chemical interactions and provided unique data on nuclear fuel behaviour under oxygen and steam starvation conditions. The individual oxidation and nitriding phenomena were well separated from each other and it was indicated by the complex microstructures in the zirconium cladding with nitrides, oxides and transition zones. The competition between different chemical reactions could be excluded in some locations and some time periods. The post-test numerical analyses of the experiment could be used for the evaluation of the role of numerical models developed for individual chemical reactions in severe accident codes.

Experimental Investigation of Natural Circulation During Air-Ingress Scenario in a Very High Temperature Reactor

In order to investigate air-ingress phenomena in a gas-cooled very high temperature reactor (VHTR), natural circulation experiments have been conducted in a helium flow loop after the injection of nitrogen into the lower plenum. A pair of helium analyzers were used to measure the nitrogen and helium concentrations in the lower plenum and upper plenum. The changes in the nitrogen concentration in the upper plenum were used to calculate the time required for the transport of nitrogen from the lower plenum to upper plenum through a riser flow channel made of graphite. The effect of system temperature and pressure on the rate of nitrogen transport has been studied extensively. Furthermore, a close examination of the graphite flow channel wall temperatures at different elevations showed small but sudden drops indicating the arrival of nitrogen at each elevation. From these data, the upward transport of nitrogen injected into the lower plenum under natural circulation conditions could be quantitatively investigated. The experimental findings indicate that the driving mechanisms for air transport through the reactor core of VHTR would result from both molecular diffusion and natural circulation. At low graphite temperatures in the riser, molecular diffusion is the dominating mechanism; however, as the riser temperature increases, natural circulation becomes dominant and the rate of nitrogen transport increases. Further, the time constants for these mechanisms have been calculated using a simplified species transport equation.

High-temperature interaction of oxygen-preloaded Zr1Nb alloy with nitrogen

Potential air ingress scenarios during accidents in nuclear reactors or spent fuel pools have raised the question of the influence of air, especially of nitrogen, on the oxidation of zirconium alloys, which are used as fuel cladding tubes and other structure materials. In this context, the reaction of zirconium with nitrogen-containing atmospheres and the formation of zirconium nitride play an important role in understanding the oxidation mechanism. This article presents the results of analysis of the interaction of the oxygen-preloaded niobium-bearing alloy M5® with nitrogen over a wide range of temperatures (800–1400°C) and oxygen contents in the metal alloy (1–7 wt.%). A strongly increasing nitriding rate with rising oxygen content in the metal was found. The highest reaction rates were measured for the saturated α-Zr(O), as it exists at the metal–oxide interface, at 1300°C. The temperature maximum of the reaction rate was approximately 100 K higher than for Zircaloy-4, already investigated in a previous study [1]. The article presents results of thermogravimetric experiments as well as posttest examinations by optical microscopy, scanning electron microscopy (SEM), and microprobe elemental analyses. Furthermore, a comparison with results obtained with Zircaloy-4 will be made.

An investigation of kinetic reaction and decomposition of sodium uranate

For the management of severe accidents of sodium-cooled fast breeder reactor, the coolability of the fuel debris bed on a core support plate is a key concern during the post-accident heat removal phase. In an air ingress scenario, the reactions between the fuel and highly oxidized sodium are likely to form sodium uranoplutonate. This would negatively influence the coolability of the fuel debris bed due to a lowering of the thermal conductivity and density. This study has focused on the formation kinetics of sodium uranate from UO2 and liquid sodium including oxygen at a high concentration. In this paper, the experiments on reaction initiation temperatures, reaction rates, and the decomposition of sodium uranate are reported.

The relationship between microstructure and oxidation effects of selected IG- and NBG-grade nuclear graphites

This study consists of three main parts. The first part characterizes IG- and NBG-grade nuclear graphites (IG-110, IG-430, NBG-18, and NBG-17) in terms of the size and shape of filler particles and how the forming method affects the pore distribution. The second part presents an experimental investigation of nuclear graphite oxidation at temperatures ranging from 700 to 1100°C in air and correlates this with the theory of active sites on graphite. Mercury porosimetry is used to quantify the pore structure development at various temperatures. X-ray diffraction analysis of selected graphites is conducted to determine the crystallographic parameters. Results of mercury porosimetry and scanning electron microscopy images are correlated with the theory of active sites on graphite in order to demonstrate the relationship between pore distribution and active sites. The third part of the study presents two experiments. The first experiment considers the effects of size of samples with the same aspect ratio and the other considers actual-sized fuel pellets and graphite sleeves to evaluate the degradation of graphite components in an air-ingress scenario.

Natural circulation patterns in the VHTR air-ingress accident and related issues

Abstract Natural circulation patterns in the VHTR during a hypothetical air-ingress accident have been investigated using computational fluid dynamic (CFD) methods in order to compare results from the previous 1-D model which was developed using GAMMA code for the air-ingress analyses. The GT-MHR 600 MWt reactor was selected to be the reference design and modeled by a half symmetric 3-D geometry using FLUENT 6.3, a commercial CFD code. CFD simulations were carried out as the steady-state calculation, and the boundary conditions were either assumed or provided from the 1-D GAMMA code results. Totally, 12 different cases have been reviewed, and many notable results have been obtained through this work. According to the simulations, natural circulation patterns in the reactor were quite different from the previous 1-D assumptions. A large re-circulation flow with thermal stratification phenomena was clearly observed in the hot-leg and the lower plenum in the 3-D model. This re-circulation flow provided about an order faster air-ingress speed (0.46 m/s in superficial velocity) than previously predicted by 1-D modeling (0.02–0.03 m/s). It indicates that the 1-D air-ingress modeling may significantly distort the air-ingress scenario and consequences. In addition, complicated natural circulation patterns are eventually expected to result in very complex graphite oxidations and corrosion behaviors.

Air-ingress analysis: Part 2—Computational fluid dynamic models

Idaho National Laboratory (INL), under the auspices of the U.S. Department of Energy (DOE), is performing research and development that focuses on key phenomena important during potential scenarios that may occur in very high-temperature reactors (VHTRs). Phenomena identification and ranking studies to date have ranked an air-ingress event, following on the heels of a VHTR depressurization, as important with regard to core safety. Consequently, the development of advanced air-ingress-related models and verification and validation data are a very high priority. Following a loss of coolant and system depressurization incident, air will enter the core of the high-temperature gas-cooled reactor through the break, possibly causing oxidation of the core and reflector graphite structure. Simple core and plant models indicate that, under certain circumstances, the oxidation may proceed at an elevated rate with additional heat generated from the oxidation reaction itself. Under postulated conditions of fluid flow and temperature, excessive degradation of lower plenum graphite because of oxidation might lead to a reactor safety issue. Computational fluid dynamics models developed in this study will improve our understanding of this phenomenon and is used to mitigate air ingress. This paper presents three-dimensional (3D) computational fluid dynamic (CFD) results for the quantitative assessment of the air-ingress phenomena. The 3D CFD simulation results show that the air-ingress accident is not controlled by molecular diffusion but density gradient driven stratified flow when the double-ended-guillotine break is assumed in a horizontal pipe configuration. It concludes that the previous air-ingress scenarios based on the molecular diffusion might not be correct and should be extensively modified to include real phenomena. This paper also presents a preliminary two-dimensional (2D) CFD simulation for validating an air-ingress mitigation concept using helium injection at the lower plenum. This simulation shows that the helium replaces air by buoyancy force and effectively mitigates air-ingress into the core.

Progress on ruthenium release and transport under air ingress conditions

A particular concern in the event of a hypothetical severe accident is the potential release of highly radiotoxic fission product (FP) isotopes of ruthenium. The highest risk for a large quantity of these isotopes to reach the containment arises from air ingress following vessel melt-through. One work package (WP) of the source term topic of the EU 6th Framework Network of Excellence project SARNET is producing and synthesizing information on ruthenium release and transport with the aim of validating or improving the corresponding modelling in the European ASTEC severe accident analysis code. The WP includes reactor scenario studies that can be used to define conditions for new experiments. The experimental database currently being reviewed includes the following programmes: • AECL experiments conducted on fission product release in air; results are relevant to CANDU loss of end-fitting accidents; • VERCORS tests on FP release and transport conducted by CEA in collaboration with IRSN and EDF; additional tests may potentially be conducted in more oxidizing conditions in the VERDON facility; • RUSET tests by AEKI investigating ruthenium transport with and without other FP simulants; • Experiments by VTT on ruthenium transport and speciation in highly oxidizing conditions. In addition to the above, at IRSN and at ENEA modelling of fission product release and of fuel oxidation is being pursued, the latter being an essential boundary condition influencing ruthenium release. Reactor scenario studies have been carried out at INR, EDF and IRSN: calculations of air ingress scenarios with respectively ICARE/CATHARE V2; SATURNE-MAAP; and ASTEC codes provided first insights of thermal-hydraulic conditions that the fuel may experience after lower head vessel failure. This paper summarizes the status of this work and plans for the future.