Amir Zacarias Mesquita1, Daniel Artur P. Palma2, Antonella Lombardi Costa3, Claubia Pereira3, Maria Auxiliadora F. Veloso3 and Patricia Amelia L. Reis3 1Centro de Desenvolvimento da Tecnologia Nuclear/Comissao Nacional de Energia Nuclear

2Comissao Nacional de Energia Nuclear 3Departamento de Engenharia Nuclear — Universidade Federal de Minas Gerais

Brazil

1. Introduction

Rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, leading to a "nuclear power renaissance" in countries the world over. In Brazil, the nuclear renaissance can be seen in the completion of construction of its third nuclear power plant and in the government’s decision to design and build the Brazilian Multipurpose research Reactor (RMB). The role of nuclear energy in Brazil is complementary to others sources. Presently two Nuclear Power Plants are in operation (Angra 1 and 2) with a total of 2000 MWe that accounts for the generation of approximately 3% of electric power consumed in Brazil. A third unity (Angra 3) is under construction. Even though with such relatively small nuclear park, Brazil has one of the biggest world nuclear resources, being the sixth natural uranium resource in the world and has a fuel cycle industry capable to provide fuel elements. Brazil has four research reactors in operation: the MB-01, a 0.1 kW critical facility; the IEA-R1, a 5 MW pool type reactor; the Argonauta, a 500 W Argonaut type reactor and the IPR-R1, a 100 kW TRIGA Mark I type reactor. They were constructed mainly for using in education, radioisotope production and nuclear research.

Understanding the behavior of the operational parameters of nuclear reactors allow the development of improved analytical models to predict the fuel temperature, and contributing to their safety. The recent natural disaster that caused damage in four reactors at the Fukushima nuclear power plant shows the importance of studies and experiments on natural convection to remove heat from the residual remaining after the shutdown. Experiments, developments and innovations used for research reactors can be later applied to larger power reactors. Their relatively low cost allows research reactors to provide an excellent testing ground for the reactors of tomorrow.

The IPR-R1 TRIGA Mark-I research reactor is located at the Nuclear Technology Development Centre — CDTN (Belo Horizonte/Brazil), a research institute of the Brazilian Nuclear Energy Commission — CNEN. The IPR-R1 reached its first criticality on November

1960 with a core configuration containing 56 aluminum clad standard TRIGA fuel elements, and a maximum thermal power of 30 kW. In order to upgrade the IPR-R1 reactor power, nine stainless steel clad fuel elements were purchased in 1971. One of these fuel elements was instrumented in the centreline with three type K thermocouples. On December 2000, four of these stainless steel clad fuel elements were placed into the core allowing to upgrading the nominal power to 250 kW. In 2004 the instrumented fuel element (IF) was inserted into all core rings and monitored the fuel temperature, allowing heat transfer investigations at several operating powers, including the maximum power of 250 kW (Mesquita, 2005). The basic safety limit for the TRIGA reactor system is the fuel temperature, both in steady-state and pulse mode operation. The time-dependence of temperature was not considered here, hence only the steady-state temperature profile was studied.

This chapter presents the experiments performed in the IPR-R1 reactor for monitoring some thermal hydraulic parameters in the fuel, pool and core coolant channels. The fuel temperature as a function of reactor power was monitored in all core rings. The radial and axial temperature profile, coolant velocity, mass flow rate and Reynolds’s number in coolant channels were monitored in all core channels. It also presents a prediction for the critical heat flux (CHF) in the fuel surface at hot channel. Data from the instrumented fuel element, pool, and bulk coolant temperature distribution were compared with the theoretical model and results from other TRIGA reactors. A data acquisition system was developed to provide a friendly interface for monitoring all operational parameters. The system performs the temperature compensation for the thermocouples. Information displayed in real-time was recorded on hard disk in a historical database (Mesquita & Souza, 2008). The data obtained during the experiments provide an excellent picture of the IPR-R1 reactor’s thermal performance. The experiments confirm the efficiency of natural circulation in removing the heat produced in the reactor core by nuclear fission (Mesquita & Rezende, 2010).

Monazite is the phosphate containing mainly ceric rare earths ((Ce, La, Nd…, Th) PO4) and is the main natural concentrator of thorium. It is widely spread (though usually in small amounts) in rocks and some types of ores. Owing to chemical and mechanical durability monazite is accumulated in placers.

The crystal structure of monazite can be presented as three-dimensional construction of oxygen nine apex polyhedron with rare-earth center atoms and oxygen tetrahedrons with the central atom of phosphorus. Nine-fold coordination allows a wide occurrence of relatively large ions of the light rare earths and thorium in mineral structure. The total content of thorium in a mineral can reach 28 wt%, and concentration of 5-7 wt% is usual. Though there are no experimental data about the form of radiogenic lead presented in the monazite structure, the numerous data, summarized for example in work [21], argued for its good stability in a monazite crystal matrix that allows monazite to be used for isotope dating.

In Ukraine monazite contains in developed fine-grained titanium-zirconium placers. By the explored easily enriched titanium-zirconium ores Ukraine comes to the forefront in Europe and in the CIS. The resources of zirconium in Ukraine make more than 10% of world ones. Now the largest Malyshevsky (Samotkansky) placer is developed and the working off of the Volchansky placer has been started.

Owing to the marked paramagnetism monazite at existing capacity of mines can be taken in passing by working out of placers in quantity of about 100 tons per year that corresponds approximately to 3.5 tons of thorium and 0.5 tons of the lead enriched with 208 isotope. Now monazite is considered as a harmful radioactive impurity and it is not produced.

The composition of monazite from the Malyshevsky placer as to the amounts of U, Th and Pb for dating purposes is well studied in work [21] by means of X-ray-fluorescent technique specially developed for individual grain analysis. In Table 3 the data about the contents of thorium, uranium and about isotope contents of lead for monazite of the Malyshevsky deposit is cited. The average composition of lead is confirmed by direct mass spectrometry determinations.

As is seen from Table 3, enrichment by 208Pb in the average for all monazite is insufficiently high. However, there is a probability of monazite separation by the flotation, magnetic or other characteristics with release of low uranium fraction of the mineral.

Extraction of total monazite concentrate by working out of the Malyshevsky placer scattering of an average almost won’t demand additional costs and its price as at first approximation can be accepted as the equal to zircon concentrate, i. e. ~ 1 US $/ kg. Cost of hydrometallurgical emanation of lead from monazite by analogy with similar processes can be estimated as (24^30 US $/ kg). The removal of differences with low U/Th ratio and the high content of 208Pb from monazite concentrate will require additional researches and will cause some rise in price of a product.

In Ukraine there are insufficiently studied shows of monazite in ancient radical breeds, their barks of aeration and in placers, i. e. enriched 208Pb. According to the available analytical data there is a possibility to detect monazite with highly enriched 208Pb.

For extraction of thorium and the lead enriched with 208 isotope Russia has a great opportunities by preparation the fine-grained titanium-zirconium placers for development and by the extraction from raw materials in complex deposits.

2. Conclusions

The paper is dedicated to the proposal of using lead enriched with the stable isotope 208Pb in FRs and ADSs instead of lead natural, natPb.

It seems that unique neutron features of 208Pb make it as one of the best among the molten metal coolants now assumed for FRs and ADSs: sodium, lead-bismuth, lead natural and others.

The main advantage of 208Pb is its low neutron absorption ability: for neutron energies En=0.1-20.0 MeV the microscopic cross sections of radiation neutron capture by 208Pb are by 1.5-2.0 times smaller as compared with natPb, and for energies, En<50 keV, the difference in the cross section values reaches 3-4 orders of magnitude. Averaged over neutron spectra of the LFR or ADS the one-group cross sections for a coolant from 208Pb are by 5-6 times smaller than those for the coolant consisted from natPb.

The second advantage of using 208Pb consists in achievement the core neutron spectra hardening on 5-6% due to low energy losses. Low neutron absorbing and moderating features of 208Pb permit to reach the gain in the multiplication factor Kef on 2-3% for critical or subcritical core fueled with U-Pu mix. In this case to have the multiplication factor Kef =1.01 for the LFR or Keff =0.97 for the ADS, both cooled with lead-208, the enrichment of power grade Pu in the U-Pu fuel can be reduced approximately on 0.7­0.8%.

The third important advantage of using 208Pb is coupled with increasing the small share of neutrons of low energies, 5-10 eV in spite of the neutron spectra hardening in whole. In this region of neutron energies the microscopic cross sections for such nuclides as 238U and 99Tc are maximum and very high, and the one-group cross sections for these nuclides averaged over neutron spectra of LFRs and ADSs cooled with lead-208 are equal to 0.6 and 0.8 barn respectively which are comparable with the one-group cross sections for typical breeders and transmutters.

The possibility of using 208Pb as coolant in commercial fast critical or subcritical reactors requires a special considering but relatively high content of this isotope in natural lead, 52.3%, and perspectives of using high performance photochemical technique of lead isotope separation permit to expect obtaining in future such a material in large quantities and under economically acceptable price. In the paper it is shown that principal possibility of acquisition of radiogenic lead containing high enriched lead -208, up to 93%, exists. Nowadays in Russian Federation and Ukraine thorium — containing loparit ores and monazite minerals are reprocessed for production of rare metal raw. Thorium and lead are not required now and they are deposited in sludge. Nevertheless, the scales of future thorium and radiogenic lead production for innovative nuclear reactors have some prospects in near-term future. The conclusion is made that to obtain the minimum amount of required in future radiogenic lead (65 t/year) for small sized FRs and ADSs the very large quantities of loparit ores or monazite minerals must be reprocessed and acquisition of radiogenic lead-208 can be economically acceptable as a co-product of rare metal raw.

The pertinent parameters required for the analysis of coolant channels are tabulated in Table 3. Figure 17 shows the power dissipate and the temperature increase in each channel at 265 kW reactor total power. This power was the results of the thermal power calibration (Mesquita et al. 2007). The profile of the mass flow rate and velocity in the core is shown in the graphs of Figure 18. Figure 19 compares experimental and theoretical profile of mass flux G in the core coolant channels. The theoretical values were calculated using PANTERA code (Veloso, 2005). As it can see by the Reynolds number the flow regime is turbulent in channels near the core centre.

Figure 20 shows the water temperatures evolution at the reactor pool, and the inlet and outlet coolant temperature in the core’s hottest channel until the beginning of steady state. The results showed that the thermocouples positioned 143 mm over the top grid plate (Inf 7) measure a temperature level higher than all the other thermocouples positioned over the reactor core. The temperature measurements above the core showed that thorough mixing of water occurs within the first centimeters above core top resulting in a uniform water temperature. It means that the chimney effect is not much high, less than 400 mm above the reactor core, in agreement with similar experiments reported by Rao et al. (1988). The chimney effect is considered as an unheated extension of the core. The chimney height is the
distance between the channel exit and the fluid isotherm plan above the core and it depends of the reactor power.

Primary Outlet

Channel

Outlet

.Channel

Fig. 20. Temperatures patterns in the reactor pool at 265 kW thermal power

1.4 Temperatures with the forced cooling system tuned off

Figure 21 shows the behavior of fuel element, channel outlet, reactor pool, and specimen rack temperatures at various operation powers, with the forced cooling system turned off.

In this chapter we have described how total absorption measurements can play an important role in improving the beta decay data necessary for summation calculations. We have discussed the technique and how its combination with IGISOL and the JYFL Penning trap has allowed us to perform measurements that had a large impact in the decay heat of 239Pu. These measurements can also be relevant for other reasons. The beta feeding distributions can also be used to deduce the beta strength and test nuclear models. This region (nuclear mass ~ 100) is interesting from the point of view of nuclear structure. For example it has been suggested that triaxial shapes play a role in this region of the nuclear chart (Moller, 2006). Most known nuclei have prolate (rugby ball shape) or spherical shapes in their ground state. If triaxiality plays a role in the structure of these nuclei, this will afect the distribution of the strength in the daughter and it may be studied using the TAS technique. Actually, we have previously used the TAS technique to infer shape effects in the A ~ 70 region (Nacher, 2004) and have started recently a related research programme in the lead region (Algora, 2005).

We plan to continue to make measurements using the TAS to obtain data of relevance to decay heat, but it is important to mention that there also similar efforts ongoing in other facilities and by other groups. In Argonne National Laboratory (Chicago, USA), there is a new facility under construction CARIBU, that will allow for the production of neutron-rich species from the fission of 252Cf. Here there are plans to use again the TAS detector employed in the measurements of Greenwood (Greenwood, 1992) and coworkers. Another example is the development of the MTAS detector by the group of Rykaczewski and coworkers, that will be used at the HRIBF facility at Oak Ridge (USA). These new facilities will contribute in the future to improving the quality of beta decay data for the decay heat application.

Our work had a large impact in 239Pu, but there is still a large amount of work to be done for 235U as was mentioned in the previous section. Additionally decays relevant for other fuels like 232Th should be also studied. Recent work by Nichols and coworkers has identified which nuclei should be measured for the 232Th fuel (Gupta, 2010).

Another aspect worth mentioning is the possible impact of these measurements in the prediction of the neutrino spectrum from reactors. In the same fashion as beta and gamma summation calculations are performed, neutrino summation calculations can be done for a working reactor. Because of the very small interaction cross section, neutrinos leave the reactor almost without interaction in the core. They carry information on the fuel composition and on the power level and their flux can not be shielded or controlled. Because of the small

interaction cross section with matter they are difficult to detect (~ 10~43cm2), but they are produced in very large numbers from the fission products. For example, approximately six antineutrinos are produced per fission, and a one GWei reactor produces of the order of 1021 neutrinos every second. The precision of the neutrino spectrum measurements can be important for neutrino oscillation experiments in fundamental physics experiments like Double CHOOZ and for non-proliferation applications (Fallot, 2007). There is presently a working group of the IAEA, which studies the feasibility of building neutrino detectors, which if positioned outside and close to a nuclear reactor can be used to monitor the power level and the fuel composition of the reactor. These measurements, if they reach the necessary precision, can be used to indicate the fuel used and to monitor manipulations of the fuel in a non-intrusive way (Porta & Fallott, 2010). We plan future measurements to address this topic of research.

Motoo Fumizawa

Shonan Institute of Technology Japan

1. Introduction

Buoyancy-driven exchange flows of helium-air were investigated through horizontal and inclined small openings. Exchange flows may occur following a window opening as ventilation, fire in the room, over the escalator in the underground shopping center as well as a pipe rupture accident in a modular high temperature gas-cooled nuclear reactor. Fuel loading pipe is located in the inclined position in the pebble bed reactor such as Modular reactor (Fumizawa, 2005, Kiso, 1999) and AVR(El-Wakil, 1982, Juni-1965, 1965).

In safety studies of High Temperature Gas-Cooled Reactor (HTGR), a failure of a standpipe at the top of the reactor vessel or a fuel loading pipe may be one of the most critical design — base accidents. Once the pipe rupture accident occurs, helium blows up through the breach immediately. After the pressure between the inside and outside of he pressure vessel has balanced, helium flows upward and air flows downward through he breach into the pressure vessel. This means that buoyancy-driven exchange flow occurs through the breach, caused by density difference of the gases in the unstably stratified field. Since an air stream corrodes graphite structures in the reactor, it is important to evaluate and reduce the air ingress flow rate during the standpipe rupture accident.

Some studies have been performed so far on the exchange flow of two fluids with different densities through vertical and inclined short tubes. Epstein(Epstein, 1988) experimentally and theoretically studied the exchange flow of water and brine through the various vertical tubes. Mercer et al. (Mercer, 1975) experimentally studied an exchange flow through inclined tubes with water and brine. He performed the experiment in the range of 3.5 <L/D < 18 and 0 deg < 0 < 90 deg, and pointed out that the length-to-diameter ratio L/D, and the inclination angle 0of the tube are the important parameter for the exchange flow rate. Most of these studies were performed on the exchange flow with a relatively small difference of the densities of the two fluids (up to 10 per cent). However, in the case of HTGR standpipe rupture accident, the density of the outside gas is at least three times larger than that of the gas inside the pressure vessel. Few studies have been performed so far in such a large range of density difference. Kang et al. (Kang, 1992) studied experimentally the exchange flow through a round tube with a partition plate. Although we may think that the partitioned plate, a kind of obstacle in the tube, decrease the exchange flow rate, he found that the exchange flow rate is increased by the partition plate because of separation of an upward and downward flow.

The objectives of the present study are to investigate the behavior of the exchange flow, i. e., exchange flow through the round long tube by several flow visualization method, then to evaluate the exchange flow rate by the PTV and PIV methods and mass increment with helium-air system. Therefore the following methods are investigated in the present study.

1. Smoke wire method

2. The optical system of the Mach-Zehnder interferometer

3. The method of the mass increment

The IPR-R1 TRIGA (Instituto de Pesquisas Radiativas — Reactor 1, Training Research Isotope production, General Atomic) is a typical TRIGA Mark I light-water and open pool type reactor. The fuel elements in the reactor core are cooled by water natural circulation. The basic parameter which allows TRIGA reactors to operate safely during either steady-state or transient conditions is the prompt negative temperature coefficient associated with the TRIGA fuel and core design. This temperature coefficient allows great freedom in steady state and transient operations. TRIGA reactors are the most widely used research reactor in the world. There is an installed base of over sixty-five facilities in twenty-four countries on five continents. General Atomics (GA), the supplier of TRIGA research reactors, since late 50’s continues to design and install TRIGA reactors around the world, and has built TRIGA reactors in a variety of configurations and capabilities, with steady state thermal power levels ranging from 100 kW to 16 MW. TRIGA reactors are used in many diverse applications, including production of radioisotopes for medicine and industry, treatment of tumors, nondestructive testing, basic research on the properties of matter, and for education and training. The TRIGA reactor is the only nuclear reactor in this category that offers true "inherent safety", rather than relying on "engineered safety". It is possible due to the unique properties of GA’s uranium-zirconium hydride fuel, which provides incomparable safety characteristics, which also permit flexibility in sitting, with minimal environmental effects (General Atomics, 2011). Figure 1 shows two photographs of the pool and the core with the IPR-R1 TRIGA reactor in operation.

The IPR-R1 TRIGA reactor core is placed at the bottom of an open tank of about 6m height and 2m diameter. The tank is filled with approximately 18 m2 of water able to assure an adequate radioactive shielding, as shown in Fig. 2. The reactor is licensed to operate at a maximum steady-state thermal power level of 100 kW, but the core and the instrumentation are configured to 250 kW, and waiting the definitive license to operate in this new power. Some of the experiments reported here were performed at power operation of 250 kW. For these experiments was obtained a provisional license for operation to this new power.

The reactor core is cooled by water natural circulation. Cooling water passage through the top plate is provided by the differential area between a triangular spacer block on top of fuel element and the round hole in the grid. A heat removal system is provided for removing heat from the reactor pool water. The water is pumped through a heat exchanger, where the
heat is transferred from the primary to the secondary loop. The secondary loop water is cooled in an external cooling tower. Figure 3 shows the forced cooling system, which transfers the heat generated in the reactor core to a water-to-water heat exchanger. The secondary cooling system transfers the reactor core heat from the heat exchanger to a cooling tower. In the diagram is shown also the instrumentation distribution and the forced and natural circulation paths in the pool.

CONTROL RODS DRIVE

ROTARY SPECIMEN RACK DRIVE MECHANISM

CENTRAL BEAM

ROTARY SPECIMEN RACK

CONCRETE

2 m

Fig. 2. IPR-R1 TRIGA reactor pool and core

A simplified view of the IPR R1 TRIGA core configuration is shown in the Fig. 4. As shown in the diagram there are small holes in the core upper grid plate. These holes were used to insert thermocouples to monitor the coolant channel temperatures. The core has a cylindrical configuration of six rings (A, B, C, D, E and F) having 1, 6, 12, 18, 24 and 30 locations respectively. These 91 positions are able to host either fuel rods or other components like control rods, a neutron source, graphite dummies (mobile reflector), irradiating and measurement channels (e. g. central thimble or A ring). Each location corresponds to a role in the aluminum upper grid plate of the reactor core. The core is surrounded by an annular graphite reflector and water. Inside the reflector there is a rotary specimen rack with 40 positions for placement of samples to be activated by neutron flux. The top view of the reactor core and the rotary specimen rack are presented in Fig. 5. There is a very high number of reactor loading configurations, so that it is possible to obtain the sub-critical level required simply loading/unloading fuel rods from the core.

The prototypical cylindrical fuel elements are a homogeneous alloy of zirconium hydride (neutron moderator) and uranium enriched at 20% in 235U. The reactor core has 58

aluminum-clad fuel elements and 5 stainless steel-clad fuel elements. One of these steel-clad fuel elements is instrumented with three thermocouples along its centreline, and was inserted in the reactor core in order to evaluate the thermal hydraulic performance of the IPR-R1 reactor (Mesquita, 2005). The fuel rod has about 3.5 cm diameter, the active length is about 37 cm closed by graphite slugs at the top and bottom ends which act as axial reflector. The moderating effects are carried out mainly by the zirconium hydride in the mixture, and on a smaller scale by light water coolant. The characteristic of the fuel elements gives a very high negative prompt temperature coefficient, is the main reason of the high inherent safety behavior of the TRIGA reactors. The power level of the reactor is controlled with three independent control rods: a Regulating rod, a Shim rod, and a Safety rod.

Start

Channel

Log v Channel

Rotary Specimen Rack

THERMOCOUPLES

Safety (C7)

NEUTRON DETECTOR

Fig. 5. Core configuration with the rotary specimen rack

A. Algora and J. L. Tain

Instituto de Fisica Corpuscular, CSIC-Univ. de Valencia, Valencia

Spain

1. Introduction

The recent incidents at the Fukushima Daiichi nuclear power plant, following the great tsunami in Japan, have shown publicly, in a dramatic way, the need for a full knowledge and proper handling of the decay heat in reactors and spent-fuel pools.

In this chapter, after a short introduction to decay heat from the historical perspective we will discuss, how the decay heat is calculated from available nuclear data, and how the quality of the available beta decay data plays a key role in the accuracy and predictive power of the calculations. We will present how conventional beta decay experiments are performed and how the deduced information from such conventional measurements can suffer from the so-called pandemonium effect. Then we will introduce the total absorption technique, a technique that can be used in beta decay experiments to avoid the pandemonium effect. Finally, we will present the impact of some recent measurements using the total absorption technique, performed by an international collaboration that we lead on decay heat summation calculations and future perspectives.

The heat generated by fission in the fuel material is conducted through the fuel, through the fuel-cladding interface, and across the cladding to the coolant. The thermal and hydrodynamic purpose of the design is to safely remove the heat generated in the fuel without producing excessive fuel temperatures or steam void formations and without closely approaching the hidrodynamic Critical Heat Flux (CHF) (Huda et al. 2001). As the IPR-R1 TRIGA reactor core power is increased, the heat transfer regime from the fuel cladding to the coolant changes from the single phase natural convection regime to subcooled nucleate boiling. The hottest temperature measured in the core channel was 65 oC (Channel 1′), below 111.4 oC, the water saturation temperature for the pressure of 1.5 bar. Therefore, the saturated nucleate boiling regime is not reached. Channel 1′ is the closest channel to the centre of the reactor where it is possible to measure the water entrance and exit temperatures. The hottest channel is Channel 0, closer to the centre. With the measured temperature values in the Channel 1′, the value of critical flow was evaluated in these two channels. The Bernath correlation was used (Eq. 5) for the calculation of the critical heat flux. With the reactor power of 265 kW operating in steady state, the core inlet temperature was 47oC. The critical flow for the Channel 0 is about 1.6 MW/m[1], giving a Departure from Nucleate Boiling Ratio (DNBR2 of 8.5. Figure 22 and Figure 23 show the values of critical flow and DNBR for the two channels. The theoretical values for reactor TRIGA of the University of New York and calculated with the PANTERA code for the IPR-R1 are also shown (General Atomic, 1970) and (Veloso, 2005). The two theoretical calculations gave smaller results than the experiments. These differences are due to the core inlet temperature used in the models.

The power reactors are projected for a minimum DNBR of 1.3. In routine operation they operated with a DNBR close to 2. The IPR-R1 reactor operates with a great margin of safety at its present power of 250 kW, the maximum heat flux in the hottest fuel is about 8 times lesser than the critical heat flux that would take the hydrodynamic crisis in the fuel cladding. This investigation indicates that the reactor would have an appropriate heat transfer if the reactor operated at a power of about 1 MW.

Isao Kataoka, Kenji Yoshida, Masanori Naitoh, Hidetoshi Okada and Tadashi Morii

Osaka University, The Institute of Applied Energy, Japan Nuclear Energy Safety Organization

Japan

1. Introduction

The accurate prediction of thermal hydraulic behavior of gas-liquid two-phase flow is quite important for the improvement of performance and safety of a nuclear reactor. In order to analyze two-phase flow phenomena, various models such as homogeneous model, slip model, drift flux model and two-fluid model have been proposed. Among these models, the two-fluid model (Ishii (1975), Delhaye (1968)) is considered the most accurate model because this model treats each phase separately considering the phase interactions at gas-liquid interfaces. Therefore, nowadays, two-fluid model is widely adopted in many best estimate codes of nuclear reactor safety. In two-fluid model, averaged conservation equations of mass, momentum and energy are formulated for each phase. The conservation equations of each phase are not independent each other and they are strongly coupled through interfacial transfer terms of mass, momentum and energy through gas-liquid interface. Interfacial transfer terms are characteristic terms in two-fluid model and are given in terms of interfacial area concentration (interfacial area per unit volume of two-phase flow) Therefore, the accurate knowledge of interfacial area concentration is quite essential to the accuracy of the prediction based on two-fluid model and a lot of experimental and analytical studies have been made on interfacial area concentration. In conventional codes based on two-fluid model, interfacial area concentration is given in constitutive equations in terms of Weber number of bubbles or droplets depending upon flow regime of two-phase flow (Ransom et al. (1985), Liles et al. (1984)). However, recently, more accurate and multidimensional predictions of two-phase flows are needed for advanced design of nuclear reactors. To meet such needs for improved prediction, it becomes necessary to give interfacial area concentration itself by solving the transport equation. Therefore, recently, intensive researches have been carried out on the models, analysis and experiments of interfacial area transport throughout the world

In view of above, in this chapter, intensive review on recent developments and present status of interfacial area concentration and its transport model will be carried out.

1.1 Experimental apparatus and procedure

The smoke wire method was used for the present investigations. Figure 1 shows a typical sketch of the apparatus. It consists of a smoke pulse generator, thin Nichrome wire with oil and a test chamber. This figure also shows the high-speed camera system, and it transfers the visual digital data to the personal computer for data acquisition. The experimental procedure is as follows. The test chamber is filled with pure helium. By removing the cover plate placed on the top of the tube, exchange flow, i. e., exchange flow of helium and air is initiated. At such condition, the smoke pulse generator ignites the high voltage. Immediately a smoke appeared and visualized the helium up flow and the air down flow in the flow path in the long tube. Test chamber diameter and height are 350 mm, the long tube on the test chamber diameter(=D) are 17.4 and 20mm, length of it (=L) are 200 and 319mm. They denote L/D=10 and L/D=18.3 respectively. It simulates a typical long tube. The inclination angle 0is 30 deg. The smoke wire conditions are as follows. The voltage is around 250 (V), current duration is 30 m sec, and the oil of thin wire is CRC-556. The high-speed camera system using D-file records the visual data up to 1600 flames in a second. Upward flows peak velocity measured by PTV method.

“S-shape smoke line

Fig. 2. Mechanism of exchange flow

(a) Ignition: 0 sec (b)Elapsed time:0.050 sec