The orientation of the heated surface significantly affects the boiling process. Boiling on a downward-facing surface is particularly challenging because bubble detachment is hindered, leading to longer bubble residence times and unique interactions than on vertical or inclined surfaces. This study investigates boiling on a large downward-facing flat surface (100 x 400 mm), focusing on critical heat flux (CHF) phenomenon. During the postulated severe accident, the situation arises in Pressurised Heavy Water Reactors (PHWRs) due to multiple failures of cooling systems and safety systems. The pressure tubes and calandria tubes have the potential to break, resulting in hot debris that falls to the bottom of the calandria vessel. The calandria vessel has a large curvature due to its larger diameter, and the bottommost portion is like a flat plate. To contain the hot debris or molten corium inside the vessel and maintain the integrity of the calandria vessel at a higher temperature is crucial to arrest the progress of a severe accident. The cooling of the vessel from outside without occurring CHF at the bottom location is important. Historically, downward-facing boiling has received limited attention, as it is normally not used in industrial applications owing to lower heat transfer and CHF values due to adverse buoyancy. Nonetheless, it is important to investigate because of the severe accident situation in PHWRs. Incorporating a simple technique of shrouds surrounding the calandria vessel can enhance the CHF by enhancing the buoyancy. This paper investigates the potential enhancement of CHF through the use of shrouds.

In advanced water-cooled nuclear reactors, the passive safety systems are preferably used to achieve enhanced safety during accidental conditions. The passive containment air cooling system (PCACS) is one of the safety systems used to remove heat from the containment during accidental conditions like a loss-of-coolant accident, station blackout, etc. in advanced nuclear reactors and small modular reactors with steel containment. The PCACS uses the buoyancy-driven flow of air to remove heat from the steel containment to avoid the over pressurization of the steel containment shell. It is extremely important to understand the natural convection around the containment shell so as to evaluate the performance of the PCACS.This paper presents an experimental investigation of the passive containment air cooling of an experimental test setup having a geometry very similar to that of an actual reactor. The air-side transient natural convection characteristics around the experimental containment system are studied in detail based on the temperature readings. The measured average heat transfer coefficient is compared with that predicted using well-known correlations available in literature. This study provides a better understanding of the natural convection flow around the containment and will help in further numerical investigations for actual-scale containments.

In some nuclear reactors, under accidental conditions, core debris forms a molten pool, which is later located in a core catcher. The core catcher proposed by the authors uses special refractory material to absorb enthalpy of corium so that temperatures are within 1500 K, which is possible to cool with side cooling and top flooding. Since performing a full-scale prototypic experiment is extremely challenging and complex because of the involvement of very high temperatures and the presence of radioactive materials, it is important to develop a Computational Fluid Dynamics (CFD) model capable of simulating coolability of the melt pool with the above cooling strategy. In the present work, a CFD model was developed for the above purpose and was benchmarked with experiments conducted under simulated conditions by the authors. The experiment involved the melting of about 25 L of sodium borosilicate glass at about 1473 K and cooling it in a scaled-down core catcher model. In the presence of decay heat inside the melt pool, turbulent natural convection plays an important role in the temperature distribution inside the melt pool and on the vessel walls. For this, we used different turbulence models. Comparisons among the Standard k-epsilon, Shear Stress Transport (SST) k-omega, and two-dimensional (2D) Large Eddy Simulation (LES) turbulence models show that SST k-omega and 2D LES turbulences are found to be in good agreement with the experimental results for the temperature distribution in the melt pool, and SST k-omega is found to be computationally less expensive than 2D LES. In general, the CFD model is capable of simulating heat transfer with phase changes inside the heat-generating melt pool. In view of this, the model can be further extended to include cooling of the melt pool in the prototype core catcher. The evolution of crust formation has been investigated in detail using a CFD model.

A severe accident involving core melt in a nuclear reactor is a major concern especially after Fukushima. Thus, to mitigate the effects of core melt accidents, an ex-vessel core catcher is being developed for Advanced Indian Nuclear Reactors. The core catcher design envisages using special refractory material. The cooling strategy of the core catcher is one of the key components in the design of the core catcher. Performing a full-scale prototypic experiment is extremely challenging and prohibitory due to the involvement of very high temperature and presence of radioactive materials. Therefore, a computational fluid dynamics (CFD) model capable of simulating the coolability of the melt pool is important to develop. In the present work, a two-dimensional (2D) CFD model was developed to understand the heat transfer phenomenon and solidification of the heat-generating simulant melt pool. The 2D symmetry geometry of the simulated core catcher vessel was used. The CFD model considers appropriate models for melting and solidification to understand crust formation in the melt pool and the k-epsilon turbulence model to resolve turbulence inside the melt pool. A decay heat of 1 MW/m(3) was also considered inside the melt pool. The CFD simulation results were compared with the authors' experimental results. The experiment involved a scaled-down ex-vessel core catcher model (CCM) employing electrical heaters to simulate decay heat. The experiment was carried out by melting about 25 L of sodium borosilicate glass using a cold crucible induction furnace at about 1200 degrees C and cooling it in the scaled-down CCM. The scaled-down CCM was strategically cooled in three phases, namely, air cooled, indirect side cooling, and complete top flooding. To overcome the complexities of simulation of the initial melt pour condition, the CFD simulation was initialized with the temperatures just after the melt pour was completed in the experiment. Similar to the experimental conditions, the CFD simulations were carried out in three phases by changing the boundary condition. Comparison of the temperatures of the melt pool by the CFD simulations and experiments at different locations gave reasonable agreement. The evolution of crust formation, melt pool temperatures, core catcher inner wall temperatures, and heat flux distribution were investigated in detail using the CFD model.

Critical heat flux (CHF) is one of the major safety limits in the nuclear power plant operation. Determining CHF for rod bundles at all operating conditions is essential; however, relying on experiments is expensive and challenging. Applicability of the empirical correlations available in the open literature to a new rod bundle is limited. In recent years, the two-fluid Eulerian approach, coupled with the heat flux partitioning model, is widely used to predict boiling flows. Especially at high-pressure conditions, computational fluid dynamics (CFD) models performance in predicting DNB in tubes are satisfactory. This work provides an overview of DNB modeling, and a state-of-the-art CFD approach for predicting the DNB is presented in detail. The complexity and limitation of the model are described while highlighting the areas requiring further efforts for understanding the physics of the same.

Core catchers are being installed in advanced light water reactors after TMI and Fukushima. The core catchers are designed to stabilise and cool molten corium within the reactor containment for a long time, reducing the risk of core meltdown. Some core catchers use sacrificial material to lower the specific volumetric heat release of the molten corium. The low density oxidic component of molten pool migrate to the top layer of the core catcher after melt-sacrificial material interaction. The high density metallic component migrate to the bottom layer. In the event that the core catcher is top flooded, this prevents metallic components from reacting with water and generate hydrogen. Post experimental analysis of the cooled stratified melt pool was performed. The experi-mental scaled down core catcher used reactor the prototypic sacrificial material while maintaining the actual reactor flooding conditions. About 550 kg of simulant corium was melted at about 2500 degrees C using heat of thermite reaction. The core catcher vessel was partially filled with sacrificial bricks. After the test, solidified debris samples from different locations of the core catcher model were analysed for density and SEM-EDS for quanti-tative and qualitative characterisation. Post-test analysis showed the melt inversion characteristics.

Natural circulation boiling water reactors (NCBWRs) are prone to flow instabilities. Due to instabilities, heat transfer deteriorates. However, the extent to which the heat transfer coefficient (HTC) deteriorates is not well established. In this work, experiments are conducted under controlled flow oscillations to evaluate the impact of the time period, mean mass flux, and amplitude ratio on HTC. Existing correlations for HTC are compared with experimental results and are found to be not satisfactory. To improve the predictions of HTC, correction factors (CFs) are proposed for existing correlations. Further, an improved HTC correlation has been developed, incorporating new nondimensional numbers.

In instances of severe accident in a nuclear reactor, the reactor containment gets pressurized and this may lead to release of hazardous radionuclides to the environment. To prevent this, containment filtered venting system (CFVS) have been envisaged in advanced reactors. One of the efficient designs in this regards is a manifold of venturi scrubber submerged in alkaline liquid and housed in a scrubber tank. A concept design of FCVS has been made using empirical models available in literature. The sizing is based on pressure and flow conditions from the containment and radioactive inventory to be scrubbed for reactor conditions. The manifold is calculated to have 30 number of venturi scrubbers submerged in alkaline pool. An experimental setup comprising of single venturi scrubber is fabricated and detailed instrumentation is provided. The flow scaling ratio is kept as 1:30. The effect of various parameters namely, gas flow rate, upstream pressure conditions, and the submergence depth is studied on hydrodynamics and the scrubbing behavior of iodine. A new CFD model having an Eulerian-Lagrangian framework is implemented for the current analysis. This model accounts for the drag forces and inertial effect of entrained jet on the continuous gas phase. The model predicts the size, velocity, and distribution of the droplets based on its interaction with the high speed gas. The retention of iodine is calculated from the gas droplet mass transfer coefficient given by Steinberg and Treybal correlation. The model is found to predict the experimental findings accurately and hence it was further employed to assess the performance of venturi scrubber under flow conditions expected in FCVS during accidental situations. The numerical model predicted that the proposed design is adequate for advanced Indian nuclear reactor.

The main aim of the present study is to aid the selection of suitable orifice configuration for helical coil oncethrough steam generators being designed for SMRs. The selection of orifice is based on the pressure drop occurring in the orifice over the desired flow range, the minimum length required for the orifice, repeatability of the results, and ease of fabrication. In the present study, three different tube inlet orifice configurations are evaluated for a helical coiled Once Through Steam Generator. Orifice length is derived for all three configurations based on the orifice pressure drop required for avoiding flow instabilities. Orifice pressure drop required to avoid flow instabilities is estimated as factor k times the sum of two phase pressure drop and super-heating zone pressure drop of the tube side flow, where minimum value of k is taken as 2 (Han et al., 2019; Kang et al., 2007). Various power levels (10%-100%) are considered to obtain the governing length of the orifice. Pressure drop across the orifice over the flow range 10%-100% is estimated theoretically using various correlations available in the literature (Idelchik). Numerical simulations are carried out for all three configurations of the orifice. Flow through the orifice falls in the turbulent regime and hence, three different turbulent models viz. k-epsilon model, RNG k-epsilon model and Realizable k-epsilon model are compared for their capability to predict pressure drop across the orifice. Theoretical and Numerical results are verified with experimental pressure drop measurements over the flow range corresponding to a power range of 10%-100%. The effect of manufacturing tolerances on pressure drop is studied experimentally for all three configurations. Five nos. of each orifice configuration were manufactured and tested experimentally for pressure drop over 10%-100% flow range to see the effect of manufacturing tolerances. It is concluded from the study that irrespective of the orifice configuration, the minimum length required for the orifice is governed by the lowest power operation i.e, the lowest flow operation. Also, a comparison of various turbulence models viz. k-epsilon model, RNG k-epsilon model and Realizable k-epsilon model, reveals that, RNG model is best suited for orifice configurations that utilize sudden expansion and contraction for creating pressure drop. Whereas, all three turbulence models compared in the present study gives similar results for orifice configurations in which pressure drop occurs mainly due to wall friction and gradual but smooth change in flow direction. Orifice configuration-3 which utilizes a helical flow path is finally selected based on the selection criteria stated above.

In subcooled boiling flows beyond a certain heat flux, heat transfer is hampered due to a phenomenon known as Departure from Nucleate Boiling (DNB). Conducting DNB experiments at one-to-one nuclear reactor operating conditions is highly challenging and expensive. Another alternative approach is to use Look-up table data. However, its applicability is limited due to its dependence on rod bundle correction factors. In the present investigation, a state-of-the-art Eulerian-Eulerian two-fluid model coupled with an extended heat flux partitioning model is used to predict DNB in tubes and rod bundles with square and hexagonal lattices (relevant to Pressurized Water Reactors). In this approach, bubble departure characteristics are modeled using semi-mechanistic models based on force balance analysis. The predicted DNB values are compared with experimental and Look-up table data and found out to be within 1.8% to 20%.

To mitigate the consequence of severe accident involving core meltdown, many advanced reactors employ ex-vessel core catchers which stabilize and cool the corium for prolonged period by strategically flooding it. The cooling system for the core catcher is one of the key components in designing the core catcher. To understand the coolability of melt pool including the effects of decay heat, a simulated experiment was performed in a scaled down ex-vessel core catcher model (CCM) employing electrical heaters to simulate decay heat of 1 MW/m3. The experiment was carried out by melting about 25 liters of sodium borosilicate glass, as a mixture of corium and sacrificial material simulant, using cold crucible induction furnace at about 1200°C. The electrical heaters were turned on as the molten corium was poured in the CCM test vessel. K-type thermocouples were used to monitor melt pool temperature as well as the bulk water temperature at different locations with time. The results show that in presence of water outside of the CCM test vessel for the present geometry, coolability of melt pool including removal of decay heat is achievable with outside vessel temperatures not exceeding 100°C. A stable crust was observed at the top surface of the melt pool, which prevented water ingression into the molten corium.

The present study is focused on debris analysis obtained from melt coolability experiments. Melt gets fragmented upon interaction with water and get quenched while forming a porous debris bed. This debris bed is coolable by natural circulation. Quenching of the melt pool has been discussed in the central region in the test section. The agglomeration of melt debris and the formation of solid cake regions adversely affected the coolability of the porous debris bed. The characterisation study of debris is carried out. The debris size range was found from 1 to 3 cm. The porosity of the debris bed was measured about 53% using the volumetric method. The data represents a systematic study of the debris agglomeration phenomenon and debris size distribution which is necessary for the development of new ways to mitigate agglomerations of debris to enable the prediction of the debris coolability in various melt coolability scenarios.

The primary objective of core catchers is to minimize the potential risk of core meltdown by stabilization and cooling of molten corium within the reactor containment for a prolonged period by strategically cooling it. Quenching of the high-temperature melt by top flooding strongly depends on the timing of injection of the cooling water, that is, immediate flooding or delayed flooding. To understand this aspect, we conducted experiments in a simulated core catcher of an Indian advanced nuclear reactor. In the experiments, sodium borosilicate glass, as a mixture of corium and sacrificial material simulant, was melted at about 1200 degrees C and cooled by water from the side of a simulated core catcher and then flooding at the top of the melt pool. Two experiments were conducted; the first one was done with immediate flooding at the top of the melt pool; the second was performed with delayed top flooding of 45 min gap. The results show that immediate flooding quenches the melt pool rapidly compared to delayed flooding. A stable crust formation at the top of the melt pool and wall adhesion is observed in the delayed flooding. It was also found that no ingression of the water in the molten pool occurred with delayed flooding. On the other hand, with immediate flooding, water was found to ingress into the melt pool due to gap formations between melt and core catcher sidewall, causing the melt to break up through eruption by ingressed water. This phenomenon resulted in the formation of sand-type debris and in faster melt cooling.

In water‐cooled nuclear reactors, the maximum power which can be extracted from the core is limited by critical heat flux (CHF). CHF in the high‐quality region is known as dryout. In advanced nuclear reactors, the coolant flow occurs solely by virtue of natural circulation; however, instabilities may occur during off‐normal operations. This may lead to premature dryout due to lower coolant flow rates seen by the heater during such oscillations. This paper describes the experimental investigation on the effect of flow oscillations on the CHF with the time period of 120 s, which is observed typically in the large‐scale natural circulation system. Based on observations made with respect to temperature transient, the continuous dryout is preceded by the transient dryout for higher flow oscillations. But as flow fluctuation decreases, the transient dryout phenomenon is found to disappear. The applicability of the look‐up table to predict CHF under oscillatory flow conditions using suitable correction factors (CFs) for premature dryout has been evaluated. CFs for the CHF under oscillations suggested by previous authors have been compared. The maximum possible degradation in CHF value suggested by previous authors has been found to agree with the present experimental data. Percentage fluctuation in heat transfer coefficient (HTC) at fully developed annular flow conditions has been evaluated, and it is found that fluctuation in HTC is in phase with the fluctuation in flow.

To mitigate severe accidents in nuclear reactors, the present research sheds light on the melt-coolability behavior of corium with hypothetical experiments that have been performed at two different nozzle diameters under bottom flooding conditions. In this research, a simulant material CaO-Fe2O3 powder mixture was melted and poured into the test section that was embedded in the test facility (using a bottom pouring furnace instead of a tiltable furnace). Then, from the bottom of the melt pool, water was flooded through a nozzle at a pressure of 0.70 bar and a water flow rate of 12 liters per minute. Because of the interaction between the water and melt, the melt quenched and converted into fine porous debris, and the temperature history was recorded using 12 K-type thermocouples connected to a data acquisition system. The average quenching time and porosity of the debris were affected by variations in the nozzle diameter. This research will help in understanding real core-melt accidents that generally occur in nuclear power plants.

Abstract Concentration solar power (CSP) systems convert solar radiation to heat and use heat engines to convert the heat to electricity. The solar receiver over which the solar radiation is concentrated and converted to heat is the most important part of the CSP. To attain maximum efficiency, the receiver in the CSP systems needs to be coated with an efficient selective solar absorber coating. In recent years, a lot of research has been focused on solar selective coatings. This has resulted in the synthesis of novel coatings that have high thermal and chemical stability, long term durability, and excellent solar selectivity making them suitable for solar thermal applications. This report reviews various solar selective coatings based on transition metals and their compounds. Various failure mechanisms are discussed in detail along with suggested prevention methods. Several thermal stability and durability tests are reported with their benefits and limitations. The effect of long-term durability on the levelized cost of coating is also discussed. Finally, we list some excellent systems and explore different ways of improving the thermal stability for SSCs, thus providing a reference for the design and optimization of new SSCs.