My Research

National Journal, Tri Dasa Mega Vol. 3 October, 2010


Mulya Juarsa1,2, Raldi Artono Koestor1, Nandy Setiadi Djaya Putra1 Anhar Riza Antariksawan2, Cukup Mulyana3, Riska Khalisa3

1Departemen Teknik Mesin, Fakultas Teknik, University of Indonesia, Depok 16424

Tel : (021) 7270011 ext 51. Fax : (021) 7270077 E-mail :

2Badan Tenaga Nuklir Nasional (BATAN), PUSPIPTEK Serpong, Tangerang Selatan 15310, Tel : (021) 7560912 Fax : (021) 7560913 E-mail :

3KBK Fisika Energi, Jurusan Fisika FMIPA, Universitas Padjadjaran Jl. Raya Bandung – Sumedang Km.21 Sumedang 45363 Tel: (022) 7796014 Fax: (022) 7792435


HEAT FLUX CALCULATION FOR BOILING CURVE DURING QUENCHING EXPERIMENT USING HEATED HOLLOW CYLINDER. One of the safety management aspects in the operation of nuclear reactors are thermal management. The basic concept of thermal management is to control the excess heat during an accident. The understanding and investigation of boiling phenomena become important research stage. Experimental study of boiling heat transfer was conducted by cooling the hollow cylinder into the water with saturation temperature. Quenching process is an important part during the occurrence of sudden cooling on heated object. A Hollow cylindrical geometry is a simulation of debris on the vertical form was used as a heated object. Cooling method is naturally hollow cylinder quenching with various initial temperatures between 300oC until 800oC into saturation temperature of water. The results as transient data was used to calculate heat flux. Film boiling regime was beginning on the temperature of the hollow cylinder reaches 300oC. The phenomenon of pool boiling was mapped in boiling curve. Whereas a pictures of the results captured by high speed camera (HSC).

Keywords: quenching, cylinder, boiling, heat flux

International Conference on Cooling and Heating Technologies (ICCHT 2010)

Boiling Mapping for Cooling Process in Annulus Narrow Gap Based on Initial Temperature Variation As a Simulation of Debris Cooling

Mulya Juarsa1,2,  Raldi A. Koestoer1,  Nandy Putra1, Anhar R. Antariksawan2, Indarto3, Bambang Riyono4

1Mechanical Engineering Department, Faculty of Engineering, Indonesia University, Depok, Indonesia,

2Thermal-Hydraulics Experimental Laboratory, Center for Reactor Technology and Nuclear Safety, National Nuclear Energy Agency (BATAN),Tangerang, Indonesia,

3Mechanical Engineering and Industry Department,Faculty of Engineering  Gadjah Mada University, Yogyakarta, Indonesia,4Indonesia-Nuclear Energy Regulatory Agency (BAPETEN), Jakarta, Indonesia


An experimental set-up which used a specific of the length of heated rod including the thermocouples for instantaneous measurement of surface temperature, was designed, constructed and used to study boiling phenomena in annulus narrow gap. Experimental was aimed to make boiling mapping during cooling process which involving heat flux. The experiments were conducted using heated rod with 700 mm length as a debris simulation and quartz glass tube to visualization purposes, the gap size between outer heated rod and inner quartz glass tube is 1.0 mm. Water with temperature of 98oC was poured into heated gap with initial variation of heated rod, respectively 1500C, 2500C, 3500C, 4500C, 5500C, and 6500C. Visualization process was conducted using high speed camera with 1000 fps. Transient temperature data was used to calculate heat flux versus wall superheat. Rewetting velocity is affected by initial temperature changes. The analysis showed the restriction from initial temperature increasing become longer, rewetting velocity in respectively 0.0072 m/s, 0.0042 m/s, 0.0015 m/s, 0.0013 m/s, 0.0010 m/s and 0.0009 m/s. The value of critical heat flux (CHF) from boiling mapping divided into two categories, lower-wall superheat for initial temperature 150oC and 250oC with CHF in respectively around 100 kW/m2 and 200 kW/m2. High-wall superheat for initial temperature 350oC, 450oC, 550oC, and 650oC by the average value of CHF around 400 kW/m2.

Keywords: boiling, CHF, rewetting, narrow gap

First Winner on The Best Paper Award of  Atom Indonesia Int. Journal in 2010


Mulya Juarsa1,2, Raldi A. Koestoer1, Nandy Putra1, Shinta Habsari4,Indarto3

1Departemen of Mechanical Engineering, Faculty of Engineering, Indonesia University,
2Center for Reactor Technology and Nuclear Safety, National Nuclear Energy Agency (BATAN),
3Mechanical Engineering and Industry Department,Faculty of Engineering Gadjah Mada University,
4National Nuclear Regulatory Agency (BAPETEN)


EXPERIMENTAL STUDY ON THE EFFECT OF INITIAL TEMPERATURE ON CHF IN A VERTICAL NARROW CHANNEL WITH BILATERAL HEATED. Study to reach the understanding of the complexities of boiling in the narrow channel which was occured in a severe accident on nuclear power plant has been carried out experimentally in order to achieve the safety management ability, CHF is one important parameter to control heat during transient accident. The methodology of research is an experiment using HeaTiNG-01 test section with modifications in the outside pipe using stainless steel material as the reactor vessel wall simulation. Experiments were conducted by heating the heated rod as a simulation of debris until the desired initial temperature by bilateral heated. Then water with a saturation temperature in atmospheric was poured gravitationally into the narrow channel. Data acquisition system recorded temperature changes in transient during the cooling process. The transient temperature profile in double heating surface and rewetting point (rewet fronts) was characterized. Experiment was conducted at three initial temperature variations i.e. 650oC, 750oC and 850oC and using channel width 1 mm. Experiment data was used to calculate heat flux then to fitting CHF form boiling curve. The results showed that CHF in outer pipe is higher than heated rod, these conditions explain that more heat is released through the outer pipe, so that the heat control can be done from outside the system to reduce the temperature quickly. In respectively base on initial temperature increase. The average value of CHF for each vertical position 100 mm and 400 mm at outer pipe are 380 kW/m2 and 733 kW/m2, and then at the heated rod are 250 kW/m2 and 497 kW/m2.

Keywords: CHF, temperature, boiling curve, bilateral

JURNAL yang akan dipublish:


Mulya Juarsa(a,b), Raldi A. Koestoer (a), Anhar R. Antariksawan (b)
(a)Mechanical Engineering Department Faculty of Engineering Indonesia University
(b)Thermal-Hydraulics Experimental Laboratory, Center for Reactor Technology and Nuclear Safety, Indonesian – National Nuclear Energy Agency

Study to understand boiling phenomena in bottom rewetting hot surface is an important step to analyze boiling heat transfer during flooding process in a simulation of Post-LOCA event. Experimental apparatus, QUEEN-II test section was designed, constructed and tested for quenching research on boiling heat transfer in bottom reflooding. Experiment has been done by heated-up SS316 rod up to 850oC, and then cooling down by water with water temperature of 90oC flowing from the bottom. The results of experiment showed that boiling regimes is quite different compare previous experiment using QUEEN-I test section. Three different values of critical heat flux (CHF) base on boiling curve are 53.51 kW/m2, 58.45 kW/m2 and 67.31 kW/m2 in respectively indicating three differences of water mass flow rate in respectively, 0.015 kg/s, 0.060 kg/s and 0.140 kg/s. The effect of 57% water mass flow rate increasing only increase 21% of CHF.

Keywords: boiling, reflooding, critical heat flux


Study on boiling heat transfer during bottom core reflooding, especially in PWR is an attractive research on nuclear engineering. This work relates to core cooling process using ECCS on Loss of Coolant Accident (LOCA). Then, the accident management must be placed to terminate the accident. This situation would be worst resulting on core melt down caused by anomaly on boiling heat transfer during reflooding process. In nuclear reactor accident, wetting hot cladding wall during ECCS injection has been studying since more than two decades, using experimental model or analytical model. X.C. Huang et al. [1] have been doing a quenching experiment using cylinder copper at pressure range between 1-10 bars and mass flux variations begin from 25 kg.m-2.s-1 until 150 kg.m-2.s-1. He analyzed boiling curve from heated rod temperature’s data. L. Spood et al. [2] observed characteristics of temperature transient at heated rod simulating PWR fuel rod. W.J. Green and K.R. Lawther [3] using ACTOR Freon loop investigated transient heat transfer at low temperature on flow boiling regime. P.K. Das et al.[4] conducted some experiments to investigate rewetting phenomena on hot vertical annular channel, his research have a good result to be use in rewetting analysis purposes, even though it was not similar. We have studied a preliminarily study using QUEEN-I test section to investigate heat transfer between heated rod outer surface and water during boiling process, but only for initial temperature 400oC, 500oC, and 600oC has been carried out[5].

The thermal-hydraulics behavior of hot vertical rod cooling during bottom flooding process and heat transfer mechanisms encountered could be flow direction-dependent. Following the blow-down phase of LOCA, the clad temperature may rise quickly to a high value (around 930oC at PWR) [6], so that the injected ECCS may not wet the clad immediately on coming into contact. Rewetting the clad is essential for effective heat removal by the emergency coolant. Extensive studies on the rewetting of hot surface have been carried out. During cooling process for a given initial temperature along rod length, the rewetting velocity may be slow down for higher initial temperature. Thus, water mass flow rate also become one of the parameters to consider in the study on boiling phenomena for a given set of initial temperature condition. Further, a majority of the earlier studies on bottom flooding including rewetting study was covered a narrow range of coolant flow rates and initial surface temperature. Therefore, study on boiling heat transfer phenomena in high temperature (850oC) during bottom reflooding simulation experiment become an important work. The present work is aimed to investigate boiling phenomena and heat transfer mechanisms using the value of CHF and related to mass flow rate in single rod using QUEEN-II test section which was connected to BETA thermal-hydraulics test loop. Temperature and mass flow rate as parameters which involves on quenching event during Post-LOCA will be investigated.

In comparison to others correlation of CHF, Monde et al. correlation [7] will be use. Monde et al. performed an extensive study of and developed a generalized correlation for CHF in asymmetrically heated vertical parallel plate channels. CHF data was obtained for water, ethanol, R113, and benzene in 10 mm deep rectangular channels formed by a copper heater and an opposing glass plate. Channel lengths of 20, 35, and 50 mm and spacings in the range of 0.45–7.0 mm were investigated, providing channel aspect ratios, L/δ, from 3 to 120.
In this experiment, the ratio of channel aspect is 45, in order to 900 mm length and gap size 20 mm. Heat flux of boiling curve also compared with other correlations. Bromley correlation [8] was used for film boiling area. Bromley was conducting a pool boiling experiment with water.
where L [m] is the characteristic length. The laminar vapor flow (LVP) with Nu = 4-5 for narrow gap was used to ensure that this cases is not vapor-water counter flow situation.

  1. Huang XC et al. Quenching experiments with a circular test section of medium thermal capacity under forced convection of water, Inter J Heat Mass Trans. 1994; 37(5): 803-818.
  2. Sepold L et el. Reflooding experiments with LWR-type fuel rod simulators in the Quench facility, Nuc Eng Des. 2001; 204: 205-220.
  3. Green WJ and Laether KR. An investigation of transient heat transfer in the region of flow boiling dryout with freon-12 in a heated tube, Nuc Eng Des. 1979; 55:131-144.
  4. Das PK. Boiling heat transfer from a single fuel pin simulator during rewetting by bottom flooding. Proceedings of ICONE 9. Nice-France; April 8 – 12, 2001.
  5. Juarsa M et al. Study on boiling heat transfer during reflooding process in “QUEEN” test section, Proceeding of ICAPP 2005. Seoul-Korea; May 15-19, 2005.
  6. Agency of Natural Resources and Energy, MITI-JAPAN. Hopes to Make Safe More Secured. How the Safety of NPP is Secured in Policy Terms, Serial Publication of NPP Safety Demonstration /Analysis.Tokyo-Japan; 2001.
  7. Monde M, Kusuda H, Uehara H. Critical heat flux during natural convective boiling in vertical rectangular channels submerged in saturated liquid, ASME J Heat Trans. 1982; 104: 300–303.
  8. Bromley LA. Heat transfer is stable film boiling, Chemical Engineering Progr. 1950; 46:221

2 responses

12 01 2012
3lelaki | Roadless traveler

c’est super. pencapaiannya sudah luar biasa pak.

12 01 2012

Terimakasih Bro….masih jauh dari yg diinginkan Bro.

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