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All Journal Atom Indonesia Journal Jurnal Teknologi Reaktor Nuklir Tri Dasa Mega Jurnal Pengembangan Energi Nuklir

Julwan Hendry Purba

Pusat Teknologi dan Keselamatan Reaktor Nuklir - BATAN

Author-ID : 431769

Decision Sciences, Operations Research & Management Earth & Planetary Sciences Electrical & Electronics Engineering Energy Engineering Environmental Science Industrial & Manufacturing Engineering Materials Science & Nanotechnology Social Sciences

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An Experimental Analysis on Nusselt Number of Natural Circulation Flow in Transient Condition Based on the Height Differences between Heater and Cooler M. Juarsa; J.P. Witoko; G Giarno; D. Haryanto; J.H. Purba
Atom Indonesia Vol 44, No 3 (2018): December 2018
Publisher : PPIKSN-BATAN

A better understanding on the phenomenon of natural circulation flow for cooling systems is necessary prior to improving the safety of nuclear power plant, not only in normal operation but also in accident conditions. One way to understand this phenomenon is by analyzing the Nusselt number in various geometrical dimensions through experimentation. The purpose of this study is to understand natural circulation phenomenon in transient condition by varying height differences between heater and cooler. To achieve this purpose, an experiment apparatus called NC-Queen was developed and arranged to enable three variations of height differences between heater and cooler, i.e., 1.4 m, 1.0 m, and 0.3 m. It is made of a stainless steel tube with a diameter of 1 inch, arranged in rectangular shape 6.4 m in length, and uses water as coolant. The initial temperature of the heater was set at 90 °C. The Nusselt number was obtained by calculating the flow rate as a function of transient temperature. The results confirm that height differences affect thermal properties and flow region based kinetics characteristics of water. In initial condition, decreasing height difference from 1.4 m to 1.0 m resulted in flow rate reduction of 16.7 %, while decreasing height difference from 1.4 m to 0.3 m resulted in a 39.1 % flow rate reduction. In final condition, the flow rate reductions were 75 % and 82.6 %, respectively. Meanwhile, in initial condition, the Nusselt number for height difference reduction from 1.4 m to 1.0 m and from 1.4 m to 0.3 m decreased by 30.5 % and 74.6 %, respectively, while for final condition, the Nusselt number decreased by 11.9 % and 67.4 %, respectively. The new constants in relationship between Nusselt number and the height difference are a = 20.06 and b = 0.56. The dominance of turbulent flow provides a good safety margin with indications of the large amount of heat released.

Preliminary Study on Mass Flow Rate in Passive Cooling Experimental Simulation During Transient Using NC-Queen Apparatus M. Juarsa; J.H. Purba; H.M. Kusuma; T. Setiadipura; S. Widodo
Atom Indonesia Vol 40, No 3 (2014): December 2014
Publisher : PPIKSN-BATAN

The research related to thermal management has been significantly inreased, especially for NPP safety. The use of passive cooling systems both during the accident and operation become reliable in the advanced reactor safety systems. Therefore it should be enhanced through experimental studies to investigate heat transfer phenomenon of the heat decay in transient cooling condition.An investigation has been performed through experiment using an NC-Queen apparatusconstructed with rectangular loop. Piping were consisting of tubes of SS316L with diameter, length, and width of 3/4 inch, 2.7 m, and 0.5 m respectively. The height between heater and cooler was 1.4 m. The experiment used initial water temperature at 70oC, 80oC, and 90oC in heater area. Transient temperature was used as experimental data to calculate water mass flow rate. The results showed that the temperature in heater area and cooler area were decreasing of about 90.6% and 95.7% at initial temperatur of 80oC, and of about 71.1% and 59.4% at initial temperature of 70oC. Those results were at higher initial temperature of 90oC compared with the initial temperature of 90oC. The average of water mass flow rate increased 81.03% from initial temperatur of 70oC. It was shown that the averages of removed heat in every second from water due to heat loss and cooler,were 3.51 watts, 5.06 watts and 6.85 watts respectively. The initial condition of heat stored in the water was quite different, but to the cooler heat removal capacity and heat loss was almost the same.Received: 10 November 2014; Revised: 23 December 2014: 24 December 2014

Reliability Study of the AP1000 Passive Safety System by Fuzzy Approach J.H. Purba; D.T. Sony Tjahyani
Atom Indonesia Vol 40, No 2 (2014): August 2014
Publisher : PPIKSN-BATAN

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/aij.2014.271

The Westinghouse AP1000 is a new design nuclear power plant which has implemented the concept of passive system. Even though a passive system may be more reliable than an active one, the possibility of the passive system to fail still exists. In line with this possibility, generic database have been used to study the reliability of the AP1000 passive safety system. However, since the used data are not specific to the AP1000, the results of the analysis will not show its real performance. This study proposes a fuzzy reliability approach to overcome this problem. The proposed fuzzy reliability approach utilizes the concept of failure possibility to qualitatively describe basic event likely occurences and membership functions of triangular fuzzy numbers to quantitatively represent qualitative failure possibilities. A case-based experiment on reliability study of the AP1000 passive safety system involved to mitigate a large break loss of collant accident is used to validate the feasibility of the proposed approach. By comparisons, probabilities of basic events generated by the proposed approach are very close to the ones which have been used by previous reliability studies. This can be observed from the small numbers of relative errors, i.e. between 0.004125 and 0.079635. These results confirm that the fuzzy reliability approach offers a more realistic technique to study the reliability of the AP1000 passive safety system without the need to engage to precise probability distributions of its components which are currently unavailable. Received: 08 November 2013; Revised: 28 May 2014; Accepted: 02 June 2014

A Fuzzy Probability Algorithm for Evaluating the AP1000 Long Term Cooling System to Mitigate Large Break LOCA J.H. Purba
Atom Indonesia Vol 41, No 3 (2015): December 2015
Publisher : PPIKSN-BATAN

Components of nuclear power plants do not always have historical failure data to probabilistically evaluate their reliability characteristics. To overcome this drawback, an alternative approach has been proposed by involving experts to qualitatively justifybasic event likelihood occurences. However, expert judgments always involve epistemic uncertainty and this uncertainty needs to be quantified. Existing fault tree analysis quantifies uncertainty using Monte Carlo simulation, which is based on probability distributions. Since expert judgments are not described in probability distributions, Monte Carlo simulation is not appropriate for evaluating epistemic uncertainty. Therefore, a new approach needs to be developed to overcome this limitation. This study proposes a fuzzy probability algorithmtoevaluate epistemic uncertainties in fault tree analysis.In the proposed algorithm, fuzzy probabilities are used to represent epistemic uncertainties of basic events, intermediate events, and the top event. To propagate and quantify epistemic uncertainty in fault tree analysis, a fuzzy multiplication rule and a fuzzy complementation rule are applied to substitute the AND Boolean and OR Boolean gates, respectively. To see the feasibility and applicability of the proposed algorithm, a case-based experiment on uncertainty evaluation of the AP1000 long term cooling system to mitigate the large break loss of coolant accident is discussed.The result shows that the best estimate probability to describe the failure of AP1000 long term cooling system generated by the proposed algorithmis3.15×10-11, which is very closed to the reference value of 1.11×10-11.This result confirms that the proposed algorithm offers a good alternative approach to quantify uncertainties in probabilistic safety assessment by fault tree analysis.Received:22 October 2014; Revised: 24 June 2015; Accepted: 29 June 2015

Co-Authors Arya Adhyaksa Waskita D. Haryanto D.T. Sony Tjahyani D.T. Sony Tjahyani D.T. Sony Tjahyani Damianus Toersiwi Sony Tjahyani Deswandri Deswandri Deswandri Deswandri Dinnia Intaningrum G Giarno H.M. Kusuma J.P. Witoko M. Juarsa M. Juarsa Ratih Luhuring Tyas S. Widodo T. Setiadipura

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