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  4. Advanced Computational Thermal hydraulics

Beside the computational tools (for both conservative and best estimate analysis) used within the Nuclear Reactor-Design Safety services, ASCOMP offers detailed thermal-hydraulic (CFD) studies involving multiphase flow and heat transfer in the reactor core and in containment systems during normal operation or during postulated transients, or in transport or storage casks under both stationary and transient conditions. TransAT (see Products) simulates medium- to small-scale 3D, thermalhydraulic multiphase flows featuring phase change and conjugate heat transfer, using either phase-average models or more detailed interface-tracking variants. Medium-scale two-phase flow behavior analysis is conducted in specific components of the system, e.g. steam generators, flow in the cold leg, etc. The hydraulic analysis of a reactor vessel, pipe lines, and valves is also covered by the CFD simulations.

Selected examples of the TransAT applications in this area are presented below:

Pressurized Thermal Shock (PTS)

RPV integrity is of utmost importance to the safe operation of nuclear power plants, and because it is one of the few components that is impractical to replace, a validated technology is essential to ensure vessel integrity over the lifetime of the plant. PTS events offer the conditions for affecting the RPV integrity and as such they constitute a pillar in nuclear safety engineering. At ASCOMP, we focus on the thermal hydraulics of PTS using a combination of advanced turbulence and multiphase flow models.
Briefly, in PTS the injected cold water mixes with the hot fluid in the cold leg and flows towards the downcomer, poor mixing may lead to extreme thermal gradients in the structures. The figures show the water free surface deformations (obtained with Level Sets) and diffusion of the heat subsequent to coolant injection.

Subchannel flow analysis

Subcooled flow boiling in PWR’s occurs in the hot fuel assemblies when heat flux is supplied to the wall, which is initially in contact with flowing liquid. Predicting this complex phenomenon is crucial for efficient operations, safety, and development of new PWRs, and more generally in LWR’s. For instance, in U.S. PWR plants, subcooled flow boiling occurs under normal operating conditions, and determines the margin to Critical Heat Flux CHF. Subcooled flow boiling also determines the rate at which corrosion products in solution in the coolant deposit on the surface of the zirconium alloy cladding, which can lead to localized corrosion and neutronic distortions (axial offset), and ultimately cladding failure.
TransAT simulations of the flow along subchannels use systematically the IST/BMR meshing technique (left) to avoid having to spend weeks on meshing the spacers in particular. Conjugate heat transfer between the rods and the fluid is thus simple to tackle, as the fully-coupled heat transfer equations are solved simultaneously. TransAT offers a hierarchy of turbulence and multiphase flow models to tackle the problem.Schowek02

Steam injection in containment pool

A typical example in passive Boiling Water Reactor (BWR) containments is the venting of vapor and gas mixtures into pressure suppression pools to control containment pressure. From the system point of view, one is interested in finding out whether there is direct communication between the exit of the vent and the surface of the pool; in this case condensation will not take place in the pool water, something that should be avoided. For the actual vent diameters and flow rates considered, experimentation at a scale of 1:1 was too expensive to consider. So, there was a strong incentive to develop and assess computational techniques capable of providing the answer. The phenomena of interest are the growth of the bubble at the vent, its rise and eventual break-up. Predicting the break-up of the bubbles is important, since after break-up the smaller bubbles condense very rapidly.
TransAT using interface tracking is an excellent tool for the predicting the behavior of these large steam bubbles and their breakup, their rate of condensation in the pool, the currents induced in the surrounding liquid and the consequent mixing.

Hydrogen distribution in the containment

In the case of a nuclear severe accident, H2 produced by the metal-water reaction and core-concrete interaction is released in the containment in which large amounts of oxygen (O2) are present. Depending on the transient and geometrical configuration, a flammable mixture can be produced, which can deflagrate, and even lead to detonation, increasing the pressure and potentially endangering the integrity of the containment. Investigations of such scenarios have repeatedly pointed out to hydrogen combustion as one of the principal phenomena causing early containment failure. Until recently, plants have traditionally analyzed the hydrogen combustion issue with 1D models such as CONTAIN or MELCOR, and as a result, the generated hydrogen mass is assumed to be uniformly distributed inside the containment. Recently though, advanced studies have demonstrated that light gases may indeed accumulate in parts of the containment, posing serious risks of deflagration, up to detonation.
Reliable 3D simulations tools for hydrogen distribution are crucial to detect potential for hydrogen combustion and suggest mitigating measures; as shown in the example below.

Mixed convection in steam generators

Steam Generators (GS) are used for heat removing from the primary circuit to the secondary circuit in pressurized water reactors. The hot water comes from the reactor core, distributes between the U-shaped tubes of the SG, and is cooled by the main feed water that is coming from the secondary circuit. Once it is cooled, the water is pumped again to the core. An SG can contain anywhere from 3’000 to 16’000 U-shaped tubes.
Mixing in SG inlet plenum is an important issue during postulated severe accidents in PWRs, where thermal stresses in steam generator tubes can be significant. Nuclear Regulatory bodies in the US and elsewhere have implemented an action plan to assess the thermal-hydraulic conditions during a PWR severe accident, with the objective to investigate the transient conditions in the hot leg and steam generator. One aspect of this research involves using state-of-the-art CFD techniques to predict inlet plenum mixing, and the TransAT simulations presented below were produced in this context.nuclear1

Boron dilution in RPV

Boric acid is used as a neutron absorber for reactivity control in PWR’s. If the boric acid concentration in the core region is reduced (boron dilution), a power excursion with possible fuel damage might occur. Such critical incidents could potentially be caused by a small break loss of coolant accident (LOCA), inducing a dilution in the steam generators tubes by reflux condensation phenomena then followed by the restart of the natural circulation. This situation can only be mitigated by the mixing at the core inlet.TransAT enables the testing a wide range of flow conditions in PWR’s pressure vessels, including with boron injection: ranging from natural convection flow-up to forced convection flow at nominal flow rates, including flow ramps due to pump start-up. For the investigation of boron dilution transients, disturbances in the cold leg are obviously easily reproducible.
The figure below displays the boron distribution in the downcomer; 2D and 3D views. The simulation results agree pretty well with the data, in particular the boron concentration at a cross flow section located just above the perforated drum. These results were obtained using an IST/BMR grid generated within 1H, and prove that the overall IST/BMR approach can be used for practical thermal-hydraulics problems in general, and RPV flows in particular.

Innovative Gen-IV HLM reactors

Thermal hydraulic studies involving free surfaces with Heavy Liquid Metals (HLM) became important for innovative Gen-IV nuclear reactors under design. In some of the reactors, the HLM can be used as coolant, while in accelerator driven systems (ADS) it can be used both as coolant and target. In SCK-CEN Belgium the first advanced design of ADS system is currently under design, with the construction of Multi-purpose hybrid research reactor for high-tech applications (MYRRHA). TransAT is used to solve several thermal-hydraulic issues encountered in MYRRHA, more specifically in the free-surface multiphase and convective thermal-flow modelling contexts. The results of water target experiment have been used for the variation and validation of TransAT models, as shown in the figure below.

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