Air Ingress Research Articles

Problems of Autoclaved Aerated Concrete Treatment

The problems of autoclave c of cellular concrete are considered. It was shown that the key problem of autoclaving is the appearance of areas of concrete with different dark colors and a reduced strength. The main reason for their occurrence is the presence of air in the porous space of cellular concrete, which prevents heat transfer. Cold air does not allow the concrete to warm up to a temperature of 170-175 °C, which hinders the flow of physical-chemical processes of hardening. The proposed activities that improve the quality of autoclave processing and precluding the formation of «non-steamings» in the concrete: ensure the temperature inside cellular concrete mass of at least 80 °C before curing, the vacuum (P=‒0.5 Bar) before pressure rise in the autoclave, preventing the ingress of air into the autoclave at the beginning of the process.

Process of Air Ingress during a Depressurization Accident of GTHTR300

A depressurization accident is the design-basis accidents of a gas turbine high temperature reactor, GTHTR300, which is JAEA’s design and one of the Very-High-Temperature Reactors (VHTR). When a primary pipe rupture accident occurs, air is expected to enter the reactor core from the breach and oxidize in-core graphite structures. Therefore, it is important to know a mixing process of different kinds of gases in the stable and unstable density stratified fluid layer. In order to predict or analyze the air ingress phenomena during the depressurization accident, we have conducted an experiment to obtain the mixing process of two component gases and the characteristics of natural circulation. The experimental apparatus consists of a storage tank and a reverse U-shaped vertical rectangular passage. One side wall of the high temperature side vertical passage is heated and the other side wall is cooled. The other experimental apparatus consists of a cylindrical double coaxial vessel and a horizontal double coaxial pipe. The outside of the double coaxial vessel is cooled and the inside is heated. The results obtained in this study are as follows. When the primary pipe is connected at the bottom of the reactor pressure vessel, onset time of natural circulation of air is affected by not only molecular diffusion but also localized natural convection. When the wall temperature difference is large, onset time of natural circulation of air is strongly affected by natural convection rather than molecular diffusion. When the primary pipe is connected at the side of the reactor pressure vessel, air will enter the bottom space in the reactor pressure vessel by counter-current flow at the coaxial double pipe break part immediately. Afterward, air will enter the reactor core by localized natural convection and molecular diffusion.

Pressure peaking phenomenon for indoor hydrogen releases

The pressure peaking phenomenon can be observed when hydrogen is released in enclosures with vent(s). The unforeseen physical phenomena of pressure peaking has been described and explained. This phenomenon occurs for hydrogen releases in enclosures where the vent(s), volume, and leak rate are such that there will be no air ingress to the enclosure. Pressure peaking describes the physical phenomenon of a peak in the pressure transient during such a release for some release conditions in a vented enclosure. This phenomenon is pronounced only for gases lighter than air, e.g. hydrogen and helium. For particular release flow rates and vent sizes the peak can be an order of magnitude higher compared to the steady-state overpressure that is reached when the enclosure is fully filled with hydrogen over time. This finding is relevant to all hydrogen applications indoors from a fuel cell in an enclosure or laboratory scale storage up to a forklift in a warehouse. The peak magnitude depends on the release flow rate, hydrogen inventory, enclosure volume and the ventilation area, and potentially can exceed the maximum pressure which the enclosure can withstand. A look up nomogram for applicability of the developed theory that is based on vent area and leak rate has been created for sustained releases. Experimental evidence of the phenomena is described. Reduced analytical equations are presented for the case of a constant flow rate release, and the associated nomogram is presented for use by hydrogen safety engineers and regulators.

Prevention of hydrogen accumulation inside the vacuum vessel pressure suppression system of the ITER facility by means of passive auto-catalytic recombiners

Hydrogen safety is a relevant topic for both nuclear fission and fusion power plants. Hydrogen generated in the course of a severe accident may endanger the integrity of safety barriers and may result in radioactive releases. In the case of the ITER fusion facility, accident scenarios with water ingress consider the release of hydrogen into the suppression tank (ST) of the vacuum vessel pressure suppression system (VVPSS). Under the assumption of additional air ingress, the formation of flammable gas mixtures may lead to explosions and safety component failure.The installation of passive auto-catalytic recombiners (PARs) inside the ST, which are presently used as safety devices inside the containments of nuclear fission reactors, is one option under consideration to mitigate such a scenario. PARs convert hydrogen into water vapor by means of passive mechanisms and have been qualified for operation under the conditions of a nuclear power plant accident since the 1990s.In order to support on-going hydrogen safety considerations, simulations of accident scenarios using the CFD code ANSYS-CFX are foreseen. In this context, the in-house code REKO-DIREKT is coupled to CFX to simulate PAR operation. However, the operational boundary conditions for hydrogen recombination (e.g. temperature, pressure, gas mixture) of a fusion reactor scenario differ significantly from those of a fission reactor. In order to enhance the code towards realistic PAR operation, a series of experiments has been performed in the REKO-4 facility with specific focus on ITER conditions. These specifically include operation under sub-atmospheric pressure (0.2–1.0 bar), gas compositions ranging from lean to rich H2/O2 mixtures, and superposed flow conditions.The paper gives an overview of the experimental program, presents results achieved and gives an outlook on the modeling approach towards accident scenario simulation.

On Oxidation Resistance Mechanisms at 1273 K of Tungsten-Based Alloys Containing Chromium and Yttria

Tungsten (W) is currently deemed the main candidate for the plasma-facing armor material of the first wall of future fusion reactors, such as DEMO. Advantages of W include a high melting point, high thermal conductivity, low tritium retention, and low erosion yield. However, was an accident to occur, air ingress into the vacuum vessel could occur and the temperature of the first wall could reach 1200K to 1450K due to nuclear decay heat. In the absence of cooling, the temperature remains in that range for several weeks. At these temperatures, the radioactive tungsten oxidizes and then volatilizes. Smart W alloys are therefore being developed. Smart alloys are supposed to preserve properties of W during plasma operation while suppressing tungsten oxide formation in case of an accident. This study focuses on investigations of thin film smart alloys produced by magnetron sputtering. These alloys provide an idealistic system with a homogeneous distribution of the elements W, chromium (Cr), and yttrium (Y) on an atomic scale. The recommended composition is W with 12 weight % of Cr and 0.5 weight % of Y. Passivation and a suppression of WO3 sublimation is shown. For the first time, the mechanisms yielding the improved oxidation resistance are analyzed in detail. A protective Cr2O3 layer forms at the surface. The different stages of the oxidation processes up to the failure of the protective function are analyzed for the first time. Using 18O as a tracer, it is shown for the first time that the oxide growth occurs at the surface of the protective oxide. The Cr is continuously replenished from the bulk of the sample, including the Cr-rich phase which forms during exposure at 1273K. A homogenous distribution of yttria within the W-matrix, which is preserved during oxidation, is a peculiarity of the analyzed alloy. Further, an Y-enriched nucleation site is found at the interface between metal and oxide. This nucleation sites are deemed to be crucial for the improved oxidation resistance.

Carbon fiber reinforced ceramic matrix composites with an oxidation resistant boron nitride interface coating

Toughening a ceramic in a ceramic matrix composite (CMC) depends on an ability of the composite to tolerate an accumulation of matrix cracks. When the reinforcement phase is carbon fiber, these cracks leave the fiber susceptible to destructive oxidation by ingress of air during high temperature exposure. Generally, a graphitic carbon interface coating is applied to carbon fibers because it provides for a weak bond between fiber and matrix that is required to promote toughening. This investigation seeks to utilize a BN coating instead of a C coating in order to promote oxidation resistance. Like graphitic carbon, BN is soft and easily cleavable. Preliminary observations that C/BN/SiC CMC's using Toray T300 carbon fibers were highly brittle and of low strength lead to a requirement of heat treating the fibers prior to the CVD of BN for toughened composites to be fabricated. It is likely heat treating removed reactive functionalities from the fiber surface to yield a weakly adhered and compliant interface.

Oxidation resistance of bulk plasma-facing tungsten alloys

Tungsten (W) currently is the main candidate as plasma-facing armour material for the first wall of future fusion reactors, like DEMO. Advantages of W include a high melting point, high thermal conductivity, low tritium retention, and low erosion yield. However, in case of an accident, air ingress into the vacuum vessel can occur and the temperature of the first wall can reach 1200 K to 1450 K due to nuclear decay heat. In the absence of cooling, the temperature will remain in that range for several weeks. At these temperatures the radioactive tungsten oxide volatilizes. Therefore, ‘smart’ W alloys are developed that aim to preserve the properties of W during plasma operation coupled with suppressed tungsten oxide formation in case of an accident.This study focusses on oxidation studies at 1273 K of samples produced by mechanical alloying followed by field assisted sintering. In a first step the sintering is optimized for tungsten (W) – chromium (Cr) -yttrium (Y) alloys. It is shown that the best oxidation resistance is achieved with submicron grain sizes. This yields a closed, protective oxide layer. In a second step the influence of the grinding process during sample preparation is analysed. It is shown that scratches initiate failure of the protective oxide. In a third step the oxidation and sublimation is measured for weeks – for the first time the sublimation is directly measured in order to determine the potential hazard in comparison to pure W. It is shown that the oxidation is suppressed in comparison to pure W. However, sublimation at a rate of 1×10−6mgcm−2s−1 starts after a few days. Nevertheless, the progess in smart alloys is evident: sublimation is delayed by about two days and complete mechanical destruction of the first wall is avoided.

Smart first wall materials for intrinsic safety of a fusion power plant

The first wall armor of a DEMOnstration fusion power plant (DEMO) is planned to be built from tungsten. However, in case of loss-of-coolant accident with air ingress, the temperature of the first wall may exceed 1000 °C due to nuclear decay heat. At such temperatures, tungsten forms volatile radioactive oxides, which may be mobilized into the environment at a rate of 10–600 kg per hour.Advanced “smart” tungsten alloys adjust their properties to the environment: during the plasma operation, preferential sputtering will form almost pure tungsten surface facing the plasma. In case of an accident, the remaining alloying elements form a protective layer, preventing tungsten mobilization.The new smart alloys contain tungsten (W), chromium (Cr) and yttrium (Y). The first bulk smart alloys produced using field-assisted sintering technique, revealed excellent oxidation resistance for a timescale of 10–20 hours. W-Cr-Y systems underwent combined plasma and oxidation test. During plasma exposure, smart alloys demonstrated nearly the same mass loss as the reference pure tungsten samples. Subsequent oxidation confirmed superior oxidation resistance of new alloys compared to the former W-Cr-Ti systems.Experiments attaining oxidation times and plasma fluence required for DEMO, are started. First results show necessity in further improvement of W-Cr-Y alloys.

Corrosion of SiC layers on coated zirconia particles in wet atmosphere

Properties of SiC coating layer in the postulated accident condition of air and water ingress is crucial to the in-pile safety of tristructural-isotropic (TRISO) coated fuel particles of high temperature gas-cooled reactors (HTR). Particles of ZrO2 kernels coated with low-density pyrolytic carbon buffer layer, high-density pyrolytic carbon layer and SiC layer fabricated by fluidized-bed chemical vapour deposition (FBCVD) were used to simulate real fuel particles in this research. The coated particles were tested at 1673 K in a certain flowing air with different water content between 0 and 100 g/m3 for 10 h. Analysis results show that the corrosion reaction took place mainly on the outer surface of SiC layer with changing infiltration depth according to the flowing gas compositions. Steam content was controlled according to the relationship between saturated water vapour pressure and environment temperature. SiO2 layer was the main product formed on the surface with thickness increase from 1.5 µm to 3 µm along with the content of water vapour in air. Formation of two types of micro-cracks after oxidation are observed: micro-cracks on spherulitic SiO2 and penetrating micro-cracks on flat SiO2. Micro-stress generation mechanism is illustrated. Three are also some bubbles are observed either in SiO2 layer or between SiO2 and SiC layers. Results showed that no fracture occurred in the SiC layer after wet oxidation.

Preliminary phenomena identification and ranking tables on the subject of the High Temperature Gas-cooled Reactor-Pebble Bed Module thermal fluids and accident analysis

The High Temperature Gas-cooled Reactor-Pebble Bed Module (HTR-PM) is a demonstration power plant designed by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University of China. In conformance with China’s nuclear safety requirements related to the development, assessment, application and quality assurance of computer software used in the safety analysis of nuclear power plants (NPPs), a phenomena identification and ranking table (PIRT) process was conducted on the subject of the HTR-PM thermal fluids and accident analysis by an expert panel sponsored by INET, on the basis of the latest HTR-PM design scheme. The PIRT objective is to identify safety-relevant phenomena that may occur during normal operation and under accident conditions, and to assess research activities pertaining to code development and assessment, such as the code validation in INET.In the PIRT process, safety-related phenomena were identified by the expert panel for the following HTR-PM scenarios: normal operation (NO), pressurized loss of forced cooling (PLOFC), depressurized loss of forced cooling (DLOFC), air ingress following DLOFC, steam/water ingress and anticipated transients without scram (ATWS). Subsequently, the relative importance of each phenomenon was evaluated according to its effect or influence on the figures of merit established by the expert panel, such as the fuel failure fraction and the maximum fuel temperature. In addition, each identified phenomenon was assigned a knowledge level rank by the expert panel, indicating the prediction capability for this phenomenon based on existing analytical tools and test data. What is more, all of the PIRT elements were well documented by a special report, including PIRT issues, objectives, descriptions of the HTR-PM plant and accident scenarios, evaluation criteria, current knowledge base, ranking scales of relative importance and knowledge level, a set of PIRTs, and expert discussions.

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.

Self-passivating tungsten alloys of the system W-Cr-Y for high temperature applications

Self-passivating tungsten based alloys for the first wall armor of future fusion reactors are expected to provide a major safety advantage compared to pure tungsten in case of a loss-of-coolant accident with simultaneous air ingress, due to the formation of a stable protective scale at high temperatures in presence of oxygen which prevents the formation of volatile and radioactive WO3. This work analyses the oxidation and thermal shock resistance of W-Cr-Y alloys obtained by mechanical alloying followed by HIPing. Alloys with different Cr and Y contents are produced in fully dense form with nanocrystalline or ultrafine-grained microstructure and a dispersion of Y-rich oxide nanoparticles located mainly at the grain boundaries. Isothermal oxidation experiments confirm an excellent oxidation resistance due to the formation of protective oxide scales at the very surface. These layers mainly consist of Cr2O3 and mixed Y-W and Cr-W oxides. The superior oxidation resistance of these alloys is confirmed by tests simulating accident-like conditions. The thermal conductivity of these alloys at 600–1000 °C is 2–3 times higher than standard Ni-base superalloys like Inconel-718. The material also exhibits outstanding thermal shock resistance: 1000 pulses of 0.19 GW/m2 power density and 1 ms duration at 400 °C base temperature resulted in no damage, while an increased power density of 0.38 GW/m2 resulted in the formation of a crack-network and slight surface roughening. An additional thermal treatment at 1550 °C improves slightly the oxidation resistance and significantly the thermal shock resistance of the alloy.

Process of air ingress during double coaxial horizontal pipe rupture accident of VHTR

Process of air ingress during double coaxial horizontal pipe rupture accident of VHTR

Thermodynamic assessment of the rhodium-ruthenium-oxygen (Rh-Ru-O) system

Ruthenium (Ru) and rhodium (Rh) are abundant platinum-group metals formed during burn-up of nuclear fuels. Under normal operating conditions, Rh and Ru accumulate and predominantly form metallic precipitates with other fission products like Mo, Pd and Tc.In the framework of vitrification of high-level nuclear waste, these fission products are poorly soluble in molten glasses. They precipitate as metallic particles and oxide phases. Moreover, these Ru and Rh rich phases strongly depend on temperature and the oxygen fugacity of the glass melt.In case of severe accidental conditions with air ingress, oxidation of the Ru and Rh is possible. At low temperatures (T < 1422 K for rhodium sesquioxide and T < 1815 K for ruthenium dioxide), the formed oxides are relatively stable. On the other hand, at high temperatures (T > 1422 K for rhodium sesquioxide and T > 1815 K for ruthenium dioxide), they may decompose into (Rh)-FCC or (Ru)-HCP metallic phases and radiotoxic volatile gaseous species.A thermodynamic assessment of the Rh-Ru-O system will enable the prediction of: (1) the metallic and oxide phases that form during the vitrification of high-level nuclear wastes and (2) the release of volatile gaseous species during a severe accident.The Calphad method developed herein employs a thermodynamic approach in the investigation of the thermochemistry of rhodium and ruthenium at high temperatures. Current literature on the thermodynamic properties and phase diagram data enables preliminary thermodynamic assessments of the Rh-O and Ru-O systems. Additionally, select compositions in the ternary Rh-Ru-O system underwent experimental tests to complement data found in literature and to establish the phase equilibria in the ternary system.

Fully ceramic microencapsulated fuel in prismatic high temperature gas-cooled reactors: Analysis of reactor performance and safety characteristics

Advanced nuclear reactor technologies have the potential to expand the missions of nuclear energy while reducing carbon emissions. This paper presents scoping reactor physics and thermal hydraulics analysis of a high temperature gas-cooled reactor (HTGR) using the fully ceramic microencapsulated (FCM) fuel form, and demonstrates the feasibility of FCM fueled HTGRs. FCM fuel consists of tristructural isotropic (TRISO) coated fuel particles embedded in a matrix of silicon carbide (SiC). The potential advantages of FCM fuel, which uses a monolithic SiC matrix, over conventional HTGR fuels with a carbon-based matrix include: a long refueling interval; high stability of the SiC matrix under irradiation with limited swelling; high fission product retention of the fuel form, with the SiC matrix acting as an additional barrier to fission product release; and enhanced oxidation resistance during normal operation and air ingress accidents. In addition, the literature shows that the effective thermal conductivity of SiC fuel compacts and conventional HTGR compacts are expected to be similar.The key finding of this study is that FCM fuel, within the form factor of a typical General Atomics prismatic graphite block, exhibits similar fuel cycle performance to conventional HTGR fuel. The reactor cycle length, discharge burnup, and natural resource utilization are similar. However, the reduced moderation in the FCM designs considered here does marginally reduce the discharge burnup, and therefore natural resource utilization, versus the reference HTGR design.The hardened neutron flux spectrum resulting from the SiC matrix, which displaces carbon from the core, requires a slightly higher packing fraction of conventional uranium oxy-carbide (UCO) fuel kernels or the use of higher density uranium mononitride (UN)-based fuel kernels. These options will marginally increase the decay power, because they harden the neutron flux energy spectrum and increase the density of 238U in the fuel. In one case considered, this will increase the absorption of neutrons in 238U, and the resultant impact of 239Np isotope on the decay power. The Doppler coefficients normalized per total fuel heat capacity are weaker in the FCM-fueled designs than in the reference HTGR design. This impacts the energy deposition in a control rod ejection accident, and hence the design of potential transient tests of these fuel forms. In addition, analyses of loss-of-forced cooling accidents indicate that the fuel temperature during these design basis accidents are up to ∼30 °C higher with FCM fuel than with conventional HTGR fuels due to the increased decay power.

Development of melt-stretching technique for manufacturing fully-recyclable thermoplastic honeycombs with tunable cell geometries

Honeycomb sandwich panels of several cell geometries were created by stretching polycarbonate melts between opposing platens. Perforations for air ingress through one of the platens were employed to enable and direct the formation of cells within the honeycomb, demonstrating a simple means to produce complicated architectures. Platen temperature, consolidation pressure, and platen movement speeds were each investigated to establish a range of effective process parameters. Honeycomb panels were successfully produced with areal densities of 0.18gcm-2 to 0.42gcm-2 and panel thicknesses ranging from 6mm to 32mm. The cell geometries were found to be effectively modeled by Voronoi diagrams seeded by the perforations used for air ingress. This model was validated by the successful production of hexagonal-, square-, and triangular-celled honeycombs, as well as an architecture combining all three cell shapes. Analysis of several samples via computed tomography provided insight into the internal distribution of material. Out-of-plane compressive testing was used to probe the mechanical performance of the structures. Minimal variation in buckling strength was found between the different honeycomb geometries, but post-failure behavior was dependent on cell shape.

Experimental study on 1/28 scaled NGNP HTGR reactor building test facility response to depressurization event

The analysis and understanding of air ingress events are an important aspect of the design of high-temperature gas-cooled reactor (HTGR) accident scenarios. These include depressurized loss of forced cooling (D-LOFC) events that allow for the possibility of air ingress into the reactor pressure vessel as a result of a break in the helium pressure boundary, which can ultimately result in oxidation of the fuel elements and other nuclear-grade graphite components. To characterize air ingress into the vented low-pressure containment of the next generation nuclear plant HTGR during hypothetical moderate-sized D-LOFC break accidents experimentally, a 1/28 scaled simplified reactor building model was established. A non-dimensional similarity approach was employed for scaling of the experimental facility. Three experiments were designed and conducted to study the dynamic response to the accident scenario. The experimental results suggest an increment in the flow area of the check valve between the reactor cavity (CV1) and steam generator cavity (CV3). Furthermore, qualitative analysis was conducted on the experimental data.

Feasibility of extravascular pacing with a novel substernal electrode configuration: The Substernal Pacing Acute Clinical Evaluation study

Subcutaneous implantable cardioverter-defibrillators provide an alternative to transvenous defibrillation but require higher shock outputs and offer no antitachycardia pacing. The Substernal Pacing Acute Clinical Evaluation (SPACE) study evaluated the feasibility of pacing from an extravascular substernal location. The primary purpose of the SPACE study was to characterize pacing from the substernal space. Secondary objectives included evaluating extracardiac stimulation and recording electrograms. The SPACE study prospectively evaluated the feasibility of pacing with a commercially available electrophysiology catheter acutely implanted in the substernal space via minimally invasive subxiphoid access. Pacing data were collected in ≥7 vectors using constant current stimulation up to 20 mA and pulse width up to 10 ms. Catheter placement was successful in all 26 patients whounderwent the procedure, with a mean placement time of 11.7 ± 10.1 minutes. Eighteen patients (69%) had successful ventricular capture in ≥1 tested vector. The mean pacing threshold at apulse width of 10 ms was 7.3 ± 4.2 mA across all vectors (5.8 ± 4.4 V). Failed capture was generally associated with suboptimal catheter placement or presumed air ingression. A low level of extracardiac stimulation was observed in 1 patient. The mean R-wave amplitude ranged from 2.98 to 4.11 mV in the unipolar configuration and from 0.83 to 3.95 mV in the bipolar configuration. The data from the SPACE study demonstrate that pacing is feasible from the extravascular substernal location. A substernal electrode configuration has the potential to provide pacing in a future extravascular device without need for intracardiac hardware placement.

Dynamic simulation of relief line during loss of insulation vacuum of the ITER cryoline

The ITER cryoline (CL) system consists of 37 types of vacuum jacketed transfer lines which forms a complex structured network with a total length of about 5 km, spread inside the Tokamak building, on a dedicated plant bridge and in the Cryoplant building/area. One of them, the low pressure relief line (RL) recovers helium discharged from process safety relief valves of the different cryogenic users and is sent it back to the Cryoplant via heater and recovery system. The process pipe diameters of the RL vary from DN 50 to DN 200 and the length is more than 1500 m. Loss of insulation vacuum (LIV) of a CL is one of the worst scenarios apart from LIV in Auxiliary Cold Boxes (ACBs). The Torus and Cryostat CL is chosen to simulate the virtual LIV and to study the anticipated behavior of the RL. Both helium LIV (LIV due to leak in helium pipe) and air LIV (LIV due to air ingress in outer vacuum jacket of the cryoline) with and without fire) have been simulated during this study. After the brief description of the CL system, the paper will describe the EcosimPro® model prepared for the dynamic study. The paper will also describe the results like minimum temperature of RL, mass flow and maximum pressure in the RL which are essentially used to choose the type and location of safety relief devices to protect the CL process pipes.

Oxidation characteristics of lead-alloy coolants in air ingress accident

The chemical behaviors of lead (Pb) based coolants in the air ingress accident of fast reactors were investigated by means of the thermodynamic considerations and the static oxidation experiments for Pb alloys at various chemical compositions. The results of the static oxidation tests for lead-bismuth (Pb-Bi) alloys indicated that Pb was depleted from the alloy due to the preferential formation of PbO in air at 773K. Bi was not involved in this oxidation procedure. Pb-Bi oxide and Bi2O3 were formed after the enrichment of Bi in the alloys due to the Pb depletion. The thickness of the oxide layer formed on liquid LBE at static condition was approximately 1mm. The oxidation rates of the alloys were much larger than that of the steels, and became larger with higher Pb concentration in the alloys. The compatibility of Pb-Bi alloys with stainless steel was worse when the Pb concentration in the alloys became low, since the dissolution type corrosion was promoted by the Bi composition in the alloy. The Pb-Li alloys were oxidized as they formed Li2PbO3 and Li2CO3. Then, Li was depleted from the alloy. These oxides of the coolant can be generated in the low temperature region of the reactor system after the increase of oxygen concentration in the coolant system in the air ingress accident.