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KUANTIFIKASI KETIDAKPASTIAN PADA ANALISIS POHON KEGAGALAN DENGAN PENDEKATAN FUZZY Julwan Hendry Purba; D.T. Sony Tjahyani
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 16, No 1 (2014): Pebruari 2014
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (312.19 KB)

Analisis pohon kegagalan dipakai untuk mengevaluasi kinerja sistem keselamatan pembangkit listrik tenaga nuklir. Analisis ini memerlukan ketersediaan data kegagalan komponen. Karena keandalan komponen dipengaruhi oleh lingkungan kerjanya maka perlu digunakan data kegagalan komponen yang berasal dari sistem yang sedang dievaluasi. Namun kenyataannya, data ini sangat sulit diperoleh sehingga penggunaan data jenerik menjadi tak terhindarkan. Penggunaan data jenerik tentunya akan menyebabkan ketidakpastian pada hasil analisis. Simulasi Monte Carlo sering dipakai untuk mengkuantifikasi ketidakpastian ini. Namun sebenarnya metode ini kurang tepat untuk mengevaluasi ketidakpastian apabila jumlah data yang dimiliki sangat terbatas. Tujuan dari penelitian ini adalah pengembangan sebuah metode analisis pohon kegagalan baru yang menerapkan konsep fuzzy untuk kuantifikasi ketidakpastian. Dalam metode baru ini, probabilitas fuzzy dipakai untuk merepresentasikan probabilitas kejadian dasar, antara serta puncak dan hukum kombinasi fuzzy dipakai untuk mengevaluasi ketidakpastian hasil analisis. Kebolehjadian gagalnya sistem injeksi akumulator AP1000 telah dievaluasi dengan menggunakan metode baru ini dan diperoleh ketidakpastian kegagalan pada interval 8,87E-12 – 8,87E-8 dengan nilai titik tengah 8,87E-10. Hasil ini membuktikan bahwa analisis pohon kegagalan dengan pendekatan fuzzy ini layak dipakai apabila yang menjadi fokus evaluasi adalah ketidakpastian karena keterbatasan data kegagalan yang dimiliki.Kata kunci: Analisis pohon kegagalan, analisis ketidakpastian, probabilitas fuzzy, hukum kombinasi fuzzy Fault tree analysis has been applied to evaluate nuclear power plant safety systems. To perform this analysis, component reliabilities need to be provided well in advance. Since working environment can affect component reliability, it is necessary to directly collect such data from the safety system being evaluated. However, due to lack of resources, such data may be unattainable. Hence, the use of generic data cannot be avoided. Unfortunately, generic data will add uncertainty to the analysis. Monte Carlo simulation has been performed to evaluate such uncertainty. However, this method is not appropriate when components do not have probability distributions of their lifetime to failures. The aim of this study is to propose a new fault tree analysis method which implements fuzzy concepts for quantifying such uncertainty. In the proposed method, fuzzy probabilities represent basic, intermediate as well as top event probabilities and fuzzy combination rules are used to evaluate the overall uncertainty of the fault tree. The proposed method has been performed to evaluate failure probability of the AP1000 accumulator injection system and generate a probability distribution between 8.87E-12 and 8.87E-8 with the point median value of 8.87E-10. This result confirms that the proposed method is feasible to evaluate system fault tree when uncertainty raised by the lack of reliability data is the main focus of the analysis.Keywords: Fault tree analysis, uncertainty analysis, fuzzy probabilities, fuzzy combination rules

ANALISIS SKENARIO KEGAGALAN SISTEM UNTUK MENENTUKAN PROBABILITAS KECELAKAAN PARAH AP1000 D.T. Sony Tjahyani; Julwan Hendry Purba
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 16, No 3 (2014): Oktober 2014
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (547.026 KB)

Kejadian Fukushima telah menunjukkan bahwa kecelakaan parah dapat terjadi, maka dari itu sangatlah penting untuk menganalisis tingkat keselamatan pada reaktor daya. Berdasarkan rekomendasi expert mission IAEA setelah kejadian Fukushima, perlu dilakukan upaya untuk meminimalisasi terjadinya kecelakaan parah yaitu dengan melakukan proses pendinginan yang maksimal. Dalam konsep keselamatan fasilitas nuklir, khususnya reaktor daya telah diterapkan konsep keselamatan berlapis (Defence in Depth, DiD). Konsep keselamatan tersebut terdiri atas 5 level pertahanan yang bertujuan mencegah dan mengurangi lepasan produk fisi ke masyarakat dan lingkungan pada saat reaktor daya mengalami kecelakaan. Dalam reaktor telah didesain sistem atau tindakan yang mempunyai fungsi untuk mengatasi setiap level tersebut. Tujuan dari analisis ini adalah menentukan probabilitas kecelakaan parah dengan melakukan skenario kegagalan sistem dalam proses pendinginan di reaktor. Sebagai obyek analisis adalah reaktor daya AP1000, karena jenis reaktor ini sedang banyak dibangun saat ini. Skenario dilakukan dengan mengasumsikan beberapa kombinasi kegagalan sistem yang termasuk dalam DiD level 2 dan 3. Kegagalan sistem kemudian dianalisis dengan menggunakan analisis pohon kegagalan berdasarkan perangkat lunak SAPHIRE ver. 6.76. Dari analisis didapatkan probabilitas gagal dari kelompok sistem DiD level 2 dan 3 pada AP1000 masih di bawah batas kriteria dari IAEA yaitu lebih kecil dari 10-2, serta probabilitas kecelakaan parah didapatkan sebesar 6,17 x 10-10. Berdasarkan analisis ini disimpulkan bahwa AP1000 mempunyai tingkat keselamatan yang cukup tinggi, karena melalui skenario kegagalan sistem didapatkan probabilitas kecelakaan parah yang sangat kecil. ABSTRACT Fukushima accident has shown that severe accident could be occurred, therefore it is important to analyze safety level of nuclear power plants. Based on the recommendations of IAEA expert mission after the Fukushima accident, necessary effort to minimize severe accident by optimizing cooling process. On the safety concept of nuclear facility especially power reactor has been applied defence in depth (DiD) concept. These concept consists of five defense levels which is to prevent and to reduce fission product release to the public and the environment when the power reactor accident happen. On the reactor has been designed system or action that have function to overcome with each those levels. The objective of this paper is to determine severe accident probability by system failure scenario on the cooling process in the reactor. The AP1000 is chosen as the reference plant to be evaluated, because currently this reactor is being built in many countries. The scenario is carried out by combining several system failures included in DiD level 2 and 3. System failure is evaluated by fault tree analysis using SAPHIRE code version 6.76. The analysis results show that the failure probability of system in the DiD level 2 and 3 AP1000 is still below the IAEA criteria limit that is less than 10-2, as well as the probability of severe accident is 6.17 x 10-10. Based on this analysis, it can be concluded that the safety level of AP1000 is high enough, because through system failure scenario is obtained the probability of severe accident is very small.

The Evaluation of the High Temperature Gas Cooled Reactor Safety to Fulfill the Requirement of the Next Generation Nuclear Julwan Hendry Purba; Arya Adhyaksa Waskita; Damianus Toersiwi Sony Tjahyani
Jurnal Pengembangan Energi Nuklir Vol 21, No 2 (2019): Desember 2019
Publisher : Pusat Kajian Sistem Energi Nuklir, Badan Tenaga Nuklir Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/jpen.2019.21.2.5615

High temperature gas cooled reactor (HTGR) has been considered to be the most promising option to meet energy demands in the future. It has also been selected as the next generation nuclear plant. The primary safety requirement of the next generation nuclear plant design is to limit radioactive material releases to practically eliminate the need for public evacuation or sheltering beyond the exclusion area boundary. The purpose of this study is to evaluate the safety design of HTGRs in order to fulfill the requirement of the next generation nuclear plant. To achieve this objective, inherent safety features, fundamental safety functions, and confinement functions realized into the design of HTGRs are comprehensively evaluated. It is found that design provisions of HTGRs can fulfill the intention of keeping radionuclides at their original sources. The layers of the coated fuel particles are very robust to retain nuclear fission products for all foreseeable reactivity events. There will be no possibility of radioactive materials to be released even though related safety systems and operator intervention are not involved in the recovery actions. This design has complied with the requirement of the next generation nuclear plant, which is to practically eliminate the need for public evacuation or sheltering beyond the exclusion area boundary.

The Implementation of Importance Measure Approaches for Criticality Analysis in Fault Tree Analysis: A Review Julwan Hendry Purba; Deswandri Deswandri
Jurnal Pengembangan Energi Nuklir Vol 20, No 1 (2018): Juni 2018
Publisher : Pusat Kajian Sistem Energi Nuklir, Badan Tenaga Nuklir Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/jpen.2018.20.1.4257

THE IMPLEMENTATION OF IMPORTANCE MEASURE APPROACHES FOR CRITICALITY ANALYSIS IN FAULT TREE ANALYSIS: A REVIEW.Fault tree analysis (FTA) has been widely applied in nuclear power plant (NPP) probabilistic safety assessment to evaluate the reliability of a safety system. In FTA, criticality analysis is performed to identify the weakest paths in the system designs and components. For this purpose, an importance measure approach can be applied. Risk managers can apply information obtained from this analysis to improve safety by implementing risk reduction measure into the new design or build a more innovative design. Various importance measure approaches have been developed and proposed for criticality analysis in FTA. Each important measure approach offers specific purposes and advantages but has limitations. Therefore, it is necessary to understand characteristics of each approach in order to select the most appropriate approach to reach the purpose of the study. The objective of this study is to review the current implementations of importance measure approaches to rank individual basic events and/or minimal cut sets regarding their contributions to the unreliability or unavailability of NPP safety systems. This study classified importance measure approaches into two groups, i.e. probability–based importance measure approaches and fuzzy–based importance measure approaches. This study concluded that clear understanding of the purpose of the study, the type of reliability data at hands, and the uncertainty in the calculation need to be considered prior to the selection of the appropriate importance measure approach to the study of interest.

Physical Ageing of The Research Reactor Core Structural Materials Due To Neutron Irradiation Exposure: A Review Julwan Hendry Purba
Jurnal Pengembangan Energi Nuklir Vol 18, No 2 (2016): Desember 2016
Publisher : Pusat Kajian Sistem Energi Nuklir, Badan Tenaga Nuklir Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/jpen.2016.18.2.3143

A research reactor (RR) is a nuclear reactor that has function to generate and utilize neutron flux and radiation ionization for research purposes and industrial applications. More than 60% of current operating RRs have been operated for 30 years or more. As the time passes, the functional capabilities of structures, systems and components (SSCs) of those RRs deteriorate by physical ageing, which can be caused by neutron irradiation exposure such as irradiation induced dislocation and microstructural changes. To extend the lifetime and/or to avoid unplanned outages, ageing on the safety related SSCs of RRs need to be properly managed. An ageing management is a strategy to engineer, operate, maintenance, and control SSC degradation within acceptablelimits. The purpose of this study is to review physical ageing of the core structural materials of the RRs caused by neutron irradiation exposure. In order to achieve this objective, a wide range of literatures are reviewed. Comprehensive discussions on irradiation behaviors are limited only on reactor vessel and core support structure materials made from zirconium and beryllium as well as their alloys, which are widely used in RRs. It is found that the stability of the mechanical properties of zirconium and beryllium as well as their alloys was mostly affected by the neutron fluences and temperatures.

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

RISK ASSESSMENT ON THE DECOMMISSIONING STAGE OF INDONESIAN TRIGA 2000 RESEARCH REACTOR Ratih Luhuring Tyas; Deswandri Deswandri; Dinnia Intaningrum; Julwan Hendry Purba
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 24, No 2 (2022): June 2022
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2022.24.2.6632

Decommissioning is the final stage of a nuclear reactor. In preparing the decommissioning plan, one of the important elements that need to be considered is safety assessment. During decommissioning, there are many complex tasks to be done where the radiological and non-radiological hazards arise and can significantly affect not only the workers but also the general public and the environment. Indonesia has no experience with nuclear reactor decommissioning, so it is necessary to study various experiences of decommissioning activities in the world. This study proposes a framework to implement the safety assessment on the decommissioning of the TRIGA 2000 research reactor. The framework was developed on desk-based research and analysis. The proposed framework involves the facility and decommissioning activities, hazard identification, hazard analysis, hazard evaluation, hazard or risk control, and independent review.

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