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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. [ABSTRACT FROM AUTHOR]

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°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. [ABSTRACT FROM AUTHOR]

The Passive Moderator Cooling System (PMCS) of the Advanced Heavy Water Reactor is designed to remove heat from the moderator passively in case of an extended station black out condition (SBO). The hot heavy-water moderator inside the Calandria rises upward due to buoyancy, cooled in a shell and tube heat exchanger (located within the loop) and returns back to the Calandria, completing a natural circulation loop. The heat exchanger, in turn, is cooled by water from a Gravity Driven Water Pool (GDWP); forming a second loop. This coupled natural circulation loop system provides sufficient cooling to prevent the increase of moderator temperature inside the Calandria vessel beyond safe limits during SBO. The feasibility of such a system should be assessed before it is implemented in the reactor. Thus, a scaled test facility was set up to simulate the thermal hydraulic characteristics of the PMCS. A set of time varying power experiments were performed, which capture the flow initiation from rest phenomena and the multidimensional natural convection flow in a coupled natural circulation system. Next, the experimental geometry was simulated using the three-dimensional computational fluid dynamics code (OpenFoam 2.2.0), which predicted temperature and flow distribution inside the system. The CFD results agree well with the experimental data within ±6%. The flow initiation phenomena shows that the initial flow generated recirculates within the Calandria, after ~900 s the flow is able to come out of the Calandria. The experiments show a time lag between the starting of the primary and secondary loop circulation. This work demonstrates the experimental and computational capability to understand and design an effective Passive Moderator Cooling System. [ABSTRACT FROM AUTHOR]

Investigations on severe accident phenomena help us in understanding the realistic accidental phenomena for the assessment of associated risk. The societal impact of radiological leakage to the environment has demanded further robustness in the line of defence of nuclear safety. Thus, to ensure the cooling and stabilization of corium within reactor containment in case of severe accident scenarios, many new reactors have been envisaged with core catcher. In this regard, corium coolability still remains an unresolved issue in spite of several efforts being taken towards its understanding. After studying the various cooling strategies, it has been demonstrated that melt coolability using bottom flooding of water is one of the most efficient techniques so far. To study the effect of decay heat on melt pool coolability under bottom flooding condition, two experiments have been performed in this paper; one without the decay heat and the other with decay heat. The test section used for carrying out these experiments consisted of two parts viz. lower part for retaining the melt from furnace, water inlet and melt quenching, and upper part for steam expansion and its outlet. The total height of the test section was 1400 mm and was made of 33 mm thick carbon steel. Total six stainless steel nozzles of diameter 12 mm were used for injecting water at the bottom of the melt pool. The lower part was surrounded by 10 radiative heaters to simulate decay heat of 10 kW which corresponds to 0.5 MW/m 3 . The experiments showed that quenching of about 25 l of melt at initial temperature of nearly 1200 °C took only a few minutes in both the cases. The debris formed in the experiments ranged between 0.5 mm fine particles to 50 mm porous chunks. The measured porosities were nearly 51% without decay heat and nearly 67% with decay heat. This paper presents the comparative experimental study of melt coolability using bottom flooding without and with decay heat scenarios. The experimental measurements show that the quenching behaviour of melt pool is not affected by decay heat, although stabilization of debris temperatures to saturation temperature takes a bit longer duration as compared to without decay heat scenarios. [ABSTRACT FROM AUTHOR]

Many advanced designs of nuclear reactors adopt a methodology of passive safety systems in order to avoid the occurence of a severe accident. In one of such systems, the decay heat generated from a reactor is transferred by natural circulation into large pool of water called the Gravity Driven Water Pool (GDWP). Three-dimensional (3D) convection flows develop, which in turn affect the heat transfer process and hence the temperature pattern. The heat transfer process can get compromised by the possible stratification of the temperature. Further, the material of construction of GDWP may have certain temperature limitations and puts bounds on the extent of stratification. The objective of this study is to investigate the 3D natural circulation phenomenon in GDWP. The present work is in continuation of our earlier work Gandhi et al. [1,2] where the natural convection has been analyzed in 0.025 and 0.21 m 3 vessels. Now, we have reported the simulation for 9247 m 3 GDWP tank. Single phase CFD simulations using open source CFD code [OpenFOAM-1.6] have been performed for a geometry (ID = 12 m, OD = 50 m and height (H T ) = 5 m). In order to reduce the thermal stratification, various geometrical modifications have been incorporated on the heat exchanger design, such as (1) distributing the heat transfer area of heat exchanger among single, double and multiple heat sources (2) to optimize the location of heat exchanger inside the GDWP (3) provision of passive elements such as draft tubes (single or concentric multiple) around the heat source at the center, which can act as a chimney. A detailed CFD analysis of three dimensional temperature and velocity distribution in the secondary side of GDWP has confirmed the mitigation of thermal stratification phenomenon by optimizing the distribution and position of heat exchangers. Passive draft tubes also result in significant enhancement in natural circulation and hence reduction in thermal stratification. [ABSTRACT FROM AUTHOR]

• Simulation of reactor and 1/10th scaled model of Passive Moderator Cooling system. • 1D numerical simulation using RELAP5/MOD3.2 of coupled natural circulation loop. • Validation of scaling philosophy was performed for a transient scenario. • Comparison of simulation results with experimental data. • Design optimization studies of Passive Moderator Cooling system. The passive moderator cooling system (PMCS) for the Advanced Heavy Water Reactor is one of the passive systems incorporated to enhance safety of the reactor. The PMCS comprises of a coupled natural circulation loop, with a Gravity-driven-water pool as the ultimate sink and a shell and tube heat exchanger as the coupling component. Previously (Pal et al., 2016), a scaled experimental facility for PMCS was setup. In this work, validity of scaling theory was verified using 1D thermal-hydraulic code (RELAP5/MOD3.2) and simulations were compared with experimental data. Finally, PMCS was simulated for various geometrical parameters, e.g. elevation difference and pipe diameters, in an attempt to optimize the design. A good agreement with experimental data (within ± 8%), except for initial 2 h, was observed in RELAP/MOD3.2 code simulation to scaled experimental facility because of internal recirculations within Calandria being more prominent than overall circulatory flows in the passive loops. [ABSTRACT FROM AUTHOR]

Passive systems are being examined for the future Advanced Nuclear Reactor designs. One of such concepts is the Passive Moderator Cooling System (PMCS), which is designed to remove heat from the moderator in the Calandria vessel passively in case of an extended Station Black Out condition. The heated heavy-water moderator (due to heat transferred from the Main Heat Transport System (MHTS) and thermalization of neutrons and gamma from radioactive decay of fuel) rises upward due to buoyancy, gets cooled down in a heat exchanger and returns back to Calandria, completing a natural circulation loop. The natural circulation should provide sufficient cooling to prevent the increase of moderator temperature and pressure beyond safe limits. In an earlier study, a full-scale 1D transient simulation was performed for the reactor including the MHTS and the PMCS, in the event of a station blackout scenario ( Kumar et al., 2013 ). The results indicate that the systems remain within the safe limits for 7 days. However, the flow inside a geometry like Calandria is quite complex due to its large size and inner complexities of dense fuel channel matrix, which was simplified as a 1D pipe flow in the aforesaid analysis. In the current work, CFD simulations are performed to study the temperature distributions and flow distribution of moderator inside the Calandria vessel using a three-dimensional CFD code, OpenFoam 2.2.0. First, a set of steady state simulation was carried out for a band of inlet mass flow rates, which gives the minimum mass flow rate required for removing the maximum heat load, by virtue of prediction of hot spots inside the Calandria. Next, a pseudo transient simulation was carried out by taking the power and mass flow rate at different time instants to the Calandria from the 1D analysis, as an input to the 3D steady state simulations. The results indicate that temperature hotspots are located at the top center of the fuel channel matrix depending on the operating regime. Furthermore, it was observed that for a period of 7 days the temperatures are well within the safe limits. [ABSTRACT FROM AUTHOR]