CFD applications in reactor safety

In nuclear reactor safety analysis, CFD has been used to perform single phase flow analysis. It is believed that for single phase flow, CFD has reached a good level of maturity and accuracy in terms of turbulence modelling, mesh, boundary conditions and numerical schemes.

However, accurate computation of multiphase flows in reactor safety analysis using CFD still needs significant R&D efforts. Furthermore, attention should also be given to the efforts leading to code and model validation.

In water cooled reactors, CFD can be used in number of safety related applications such as given in [52] and [53] which include, among others:

(a) Boron dilution transient;

(b) Pressurized thermal shock;

(c) Hydrogen distribution;

(d) Aerosol deposition;

(e) Direct contact condensation;

(f) Natural circulation in various reactor components;

(g) Bubble dynamics in suppression pools;

(h) Flow mixing in lower plenum;

(i) Steam line break;

(j) Thermal fatigue;

Following is an explanation of some of these safety related issues, specifically (a) to (e) on the previous list, and why CFD is used to simulate them.

(a) Boron dilution transient

Boron dilution is one of the important transient in PWR. In this transient, slugs of weakly borated water can be formed in one of the primary system loops due to different mechanism such as failure of water makeup system, steam generator tube break etc. As soon as the under borated slug is entered into the reactor core, it results into insertion of positive reactivity and might lead to power excursion. The transport of under borated slug of water in to the core may lead to serious consequences with respect of reactor safety especially after RCP startup or during restart of natural circulation after LOCA.

The boron dilution transient can be assessed in three steps:

(i) Global scale system code analysis which provide boundary conditions for step (ii);

(ii) Local scale CFD analysis which provide detailed 3D simulations in cold leg, downcomer, and reactor core;

(iii) Neutronic analysis using input from step (ii).

For modelling this transient in CFD, the related parts of the reactor are downcomer, the lower plenum, and the piping network used for transporting the slug. The features relevant to CFD simulations are the geometrical complexity of the domain, transient behaviour of the flow, and the need of the precise mixing properties of the flow.

(b) Pressurized thermal shock

PTS in general denotes the occurrence of thermal loads on the reactor pressure vessel under pressurized conditions. PTS is related with the ageing of the RPV as its integrity must need to be assured throughout the lifetime of the vessel. A very severe PTS scenario limiting the RPV life is injection of cold water from emergency core cooling system into the cold leg during a postulated loss of coolant accident. The water after injection gets mixed with the hot water present in the cold leg, and the mixture then flows towards the downcomer resulting into further mixing with the ambient fluid as shown in FIG. 44. As a result, high thermal gradients may occur in the structural components while the primary circuit pressurization is partially preserved.

The main issue in simulating PTS condition is to predict the transient fluid temperature reliably to estimate the loads upon the RPV in order to investigate thermal stresses and thermal fatigue of the vessel. The temperature of the fluid is influenced by turbulence, stratification and by condensation rate in case of two phase flow. It is important to know that the state (single phase

or two phase) of the cooling fluid depends upon the break size, its location, and operational mode of the nuclear power plant.

FIG. 44. Flow feature present during PTS condition with partially filled cold leg [53].

As far as CFD simulations for PTS are concerned, it is important to capture above mentioned features for the whole transient which may last for several minutes. Furthermore, complex geometries (downcomer, upper plenum, and pipes) and complex flow patterns (such as stratified flow, jet etc.), and unsteady 3D effects make this a challenging problem for CFD. The existing reported simulations concern single phase flow, whereas simulations of two phase flows in such situations are just beginning. The detailed references on this problem can be found in [52].

(c) Hydrogen distribution

During a severe accident in nuclear power plant containment, large quantities of hydrogen are produced. If the amount of hydrogen exceeds prescribed limit, it can result in deflagration or even detonation putting the integrity of containment at stake, the last and final barrier against the spread of radioactivity. This situation requires the proper distribution and handling of hydrogen inside the containment. Historically, the severe accidents at TMI 1979, Chernobyl 1986 and recently Fukushima Daiichi 2011 all suffered with the core damage and eventually hydrogen generation. In TMI, over 500 kg of hydrogen was produced while in Chernobyl 3000 kg per unit [54]. Consequently, radioactivity was released outside the containment building resulting in the evacuation of public in the immediate periphery of nuclear power plants.

In order to employ the effective hydrogen mitigation methods, it is important to know the detailed thermal hydraulics phenomena associated with the containment. The containment related thermal hydraulics phenomena include condensation and evaporation on walls, pool

ECCS injection Momentum transfer (gas–liquid interface) Turbulence (gas–

liquid interface)

surfaces and condensers. These phenomena need to be accurately modelled as the associated heat and mass transfer strongly influence the pressure and mixture composition inside containment. The accurate prediction of this mixture composition is very important as this information is required to determine the burning mode of hydrogen and operation of passive autocatalytic recombiners (PARs) inside the containment.

Since the scale of the problem is large due to the huge size of the containment, capturing all the essential features using CFD require huge computational effort. In case of coarse mesh, accuracy is lost and all the true gradients of velocity and temperature are washed out resulting into a wrong distribution of hydrogen inside the containment. Furthermore, condensation model need to be incorporated as these are not standard to commercial CFD codes. Similarly, in order to simulate realistic accidental scenarios, there is need to run simulations with complete geometrical model including PARs and other containment related systems such as sprays and sumps.

(d) Aerosol deposition and atmospheric transport

In the case of severe accident, fission products are released inside the containment in the form of aerosols. If the containment integrity is not maintained and there is a leak in the containment, these aerosols would be released into the environment and pose a health hazard to the population around the nuclear power plant. In the conservative scenario, all the aerosols eventually are released into the environment. However, an accurate realistic prediction can be made by investigating the detailed processes which include initial core degradation, fission product release, aerosol borne transport and retention in the coolant circuitry, and the aerosol dynamics and chemical behaviour in the containment.

The phenomena of aerosol removal from the containment requires in depth treatment of the forces acting on the aerosol particles. In order to simulate and track their motion, CFD simulations are needed in order to predict their 3-D velocity field. The goal in such simulations is to find out the residence time or life span of the aerosols moving in the containment using CFD. By accurately tracking all the aerosol particles, statistical information on the actual deposition can be obtained.

(e) Direct contact condensation (DCC)

In some nuclear reactor designs, saturated steam is discharged to subcooled water pools where it gets condensed. This phenomenon is known as direct contact condensation and may occur as part of the nuclear reactor emergency core cooling system. When DCC occurs, it causes thermal and pressure oscillations, which may challenge the structural integrity of related equipment and piping. Therefore, investigations of DCC mechanisms have been performed to find the efficiency of the condensation process and thermal mixing in the pool. To perform such investigations for nuclear reactor design and safety, complete 3-D simulations using CFD would be required. However, CFD investigations related to DCC mechanisms are sparse compared to experimental investigations. Some experimental and CFD studies on DCC are given in [55] and [56].