Pub Date : 2018-09-05DOI: 10.17146/JPEN.2018.20.1.4143
Ferhat Aziz
PENELITIAN DAN PENGEMBANGAN MATERIAL STRUKTUR REAKTOR MAJU. Pengoperasian suatu pembangkit listrik tenaga nuklir secara selamat dan ekonomis sangat tergantung kepada ketersediaan material struktur yang handal. Selama umur operasi reaktor yang dewasa ini mencapai 60 tahun, material struktur terpapar temperatur yang tinggi, lingkungan korosif dan medan radiasi kuat yang terjadi akibat proses fisi. Struktur penopang bahan bakar akan memiliki umur lebih singkat dalam lingkungan demikian. Sementara itu tuntutan dalam pengembangan reaktor maju masa depan adalah temperatur operasi dan burnup yang lebih tinggi untuk meningkatkan efisiensi operasi reaktor. Material yang mampu mengatasi kondisi itu harus stabil secara dimensional dalam medan radiasi, baik dengan maupun tanpa adanya stress, serta memiliki sifat yang baik di lingkungan yang korosif. Makalah ini menelaah secara singkat litbang material struktur reaktor yang potensial digunakan dalam sistem reaktor maju, seperti baja ferritic/martensitic, baja ODS, dan keramik, termasuk sintesis baja ODS yang dikembangkan di BATAN. Selain itu dikaji pula kerusakan radiasi pada material struktur serta sintesis dan simulasi kinerja perisai radiasi menggunakan paket program MCNP5. Hasil kajian menunjukkan bahwa walaupun hasil yang ada sudah cukup baik, masih diperlukan litbang lebih lanjut guna mendapatkan material yang memenuhi kriteria yang diinginkan, baik melalui eksperimen maupun secara simulasi dan pemodelan dengan komputer.
{"title":"Penelitian dan Pengembangan Material Struktur Reaktor Maju","authors":"Ferhat Aziz","doi":"10.17146/JPEN.2018.20.1.4143","DOIUrl":"https://doi.org/10.17146/JPEN.2018.20.1.4143","url":null,"abstract":"PENELITIAN DAN PENGEMBANGAN MATERIAL STRUKTUR REAKTOR MAJU. Pengoperasian suatu pembangkit listrik tenaga nuklir secara selamat dan ekonomis sangat tergantung kepada ketersediaan material struktur yang handal. Selama umur operasi reaktor yang dewasa ini mencapai 60 tahun, material struktur terpapar temperatur yang tinggi, lingkungan korosif dan medan radiasi kuat yang terjadi akibat proses fisi. Struktur penopang bahan bakar akan memiliki umur lebih singkat dalam lingkungan demikian. Sementara itu tuntutan dalam pengembangan reaktor maju masa depan adalah temperatur operasi dan burnup yang lebih tinggi untuk meningkatkan efisiensi operasi reaktor. Material yang mampu mengatasi kondisi itu harus stabil secara dimensional dalam medan radiasi, baik dengan maupun tanpa adanya stress, serta memiliki sifat yang baik di lingkungan yang korosif. Makalah ini menelaah secara singkat litbang material struktur reaktor yang potensial digunakan dalam sistem reaktor maju, seperti baja ferritic/martensitic, baja ODS, dan keramik, termasuk sintesis baja ODS yang dikembangkan di BATAN. Selain itu dikaji pula kerusakan radiasi pada material struktur serta sintesis dan simulasi kinerja perisai radiasi menggunakan paket program MCNP5. Hasil kajian menunjukkan bahwa walaupun hasil yang ada sudah cukup baik, masih diperlukan litbang lebih lanjut guna mendapatkan material yang memenuhi kriteria yang diinginkan, baik melalui eksperimen maupun secara simulasi dan pemodelan dengan komputer.","PeriodicalId":179965,"journal":{"name":"Jurnal Pengembangan Energi Nuklir","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2018-09-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"125130451","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2018-09-05DOI: 10.17146/JPEN.2018.20.1.4314
Dedy Priambodo
PENENTUAN NILAI JARAK AMAN SUMBER TIDAK BERGERAK: SKENARIO KEBAKARAN DAN LEDAKAN PADA SPBU DAN SPPBE DI SEKITAR TAPAK RDE. Nilai jarak aman (Screening Distance Value, SDV) untuk tiap-tiap jenis sumber bahaya harus ditentukan menggunakan pendekatan konservatif sehingga pengaruh kejadian interaksi yang berada di luar jarak ini tidak perlu diperhitungkan lebih lanjut. Dari pengamatan, sumber tidak bergerak yang paling banyak berada di sekitar tapak adalah Stasiun Pengisian Bahan Bakar Umum (SPBU) dan Stasiun Pengangkutan dan Pengisian Bulk Elpiji (SPPBE). Oleh karena itu, penelitian ini bertujuan untuk menentukan SDV dari SPBU dan SPPBE untuk skenario ledakan dan kebakaran dengan lokasi studi di sekitar tapak Reaktor Daya Eksperimental (RDE), Puspiptek Serpong. Studi ini menggunkan Areal Location Of Hazardous Atmospheres (ALOHA) untuk mensimulasikan zona ancaman ledakan dan kebakaran dari substansi Bensin dan LPG. Simulasi ini melibatkan ledakan dan kebakaran iso-oktana di SPBU dan butane di SPPBE untuk menganalisis zona ancaman dari skenario kecelakaan. Diketahui bahwa SDV sangat tergantung dari jenis, maupun kapasitas material berbahaya, skenario kecelakaan serta kondisi atmosfer pada sumber bahaya tersebut. Dari penelitian ini, diketahui pula suatu sumber bahaya bisa mempunyai SDV lebih dari satu nilai. SDV SPBU berkapasitas 170.000 liter untuk skenario kebakaran 976 m dan 1.200 m, sedangkan untuk skenario ledakan adalah 958 m dan 1.200 m. SDV SPPBE berkapasitas 25.000 kg untuk skenario kebakaran 445 m dan 565 m, sedangkan untuk skenario ledakan adalah 430 m dan 573 m. Sementara SDV SPPBE berkapasitas 100.000 kg untuk skenario kebakaran 913 m dan 1.100 m, sedangkan untuk skenario ledakan adalah 758 m dan 1.100 m. Studi ini dapat memberikan alternatif penentuan nilai SDV pada studi dan evaluasi tapak reaktor nuklir.
{"title":"Penentuan Nilai Jarak Aman Sumber Tidak Bergerak: Skenario Kebakaran dan Ledakan pada SPBU dan SPPBE di Sekitar Tapak RDE","authors":"Dedy Priambodo","doi":"10.17146/JPEN.2018.20.1.4314","DOIUrl":"https://doi.org/10.17146/JPEN.2018.20.1.4314","url":null,"abstract":"PENENTUAN NILAI JARAK AMAN SUMBER TIDAK BERGERAK: SKENARIO KEBAKARAN DAN LEDAKAN PADA SPBU DAN SPPBE DI SEKITAR TAPAK RDE. Nilai jarak aman (Screening Distance Value, SDV) untuk tiap-tiap jenis sumber bahaya harus ditentukan menggunakan pendekatan konservatif sehingga pengaruh kejadian interaksi yang berada di luar jarak ini tidak perlu diperhitungkan lebih lanjut. Dari pengamatan, sumber tidak bergerak yang paling banyak berada di sekitar tapak adalah Stasiun Pengisian Bahan Bakar Umum (SPBU) dan Stasiun Pengangkutan dan Pengisian Bulk Elpiji (SPPBE). Oleh karena itu, penelitian ini bertujuan untuk menentukan SDV dari SPBU dan SPPBE untuk skenario ledakan dan kebakaran dengan lokasi studi di sekitar tapak Reaktor Daya Eksperimental (RDE), Puspiptek Serpong. Studi ini menggunkan Areal Location Of Hazardous Atmospheres (ALOHA) untuk mensimulasikan zona ancaman ledakan dan kebakaran dari substansi Bensin dan LPG. Simulasi ini melibatkan ledakan dan kebakaran iso-oktana di SPBU dan butane di SPPBE untuk menganalisis zona ancaman dari skenario kecelakaan. Diketahui bahwa SDV sangat tergantung dari jenis, maupun kapasitas material berbahaya, skenario kecelakaan serta kondisi atmosfer pada sumber bahaya tersebut. Dari penelitian ini, diketahui pula suatu sumber bahaya bisa mempunyai SDV lebih dari satu nilai. SDV SPBU berkapasitas 170.000 liter untuk skenario kebakaran 976 m dan 1.200 m, sedangkan untuk skenario ledakan adalah 958 m dan 1.200 m. SDV SPPBE berkapasitas 25.000 kg untuk skenario kebakaran 445 m dan 565 m, sedangkan untuk skenario ledakan adalah 430 m dan 573 m. Sementara SDV SPPBE berkapasitas 100.000 kg untuk skenario kebakaran 913 m dan 1.100 m, sedangkan untuk skenario ledakan adalah 758 m dan 1.100 m. Studi ini dapat memberikan alternatif penentuan nilai SDV pada studi dan evaluasi tapak reaktor nuklir.","PeriodicalId":179965,"journal":{"name":"Jurnal Pengembangan Energi Nuklir","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2018-09-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"130021950","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2018-09-05DOI: 10.17146/JPEN.2018.20.1.4083
S. M. Lumbanraja, E. Liun
REVIU IMPLEMENTASI THORCON MOLTEN SALT REACTOR DI INDONESIA. Thorcon Molten Salt Reactor (Thorcon MSR) merupakan jenis PLTN generasi ke-4 yang dirancang oleh Martingle Inc. Amerika Serikat. PLTN ini berbahan bakar, dan berpendingin garam cair , beroperasi pada temperatur tinggi dan tekanannya mendekati tekanan atmosfir. PLTN ini berdaya 1000 MWe yang dihasilkan oleh 4 modul dengan daya masing-masing 250 MWe. Makalah ini bertujuan untuk menguraikan dan memberi gambaran tentang kelayakan ThorCon MSR kepada pemangku kepentingan ketika akan diimplementasikan di Indonesia. Metodologi yang digunakan adalah mempelajari berbagai pustaka tentang teknologi Thorcon, ketersediaan SDM, partisipasi nasional dan regulasinya, serta dilakukan dengan analisis SWOT (Strength, Weaknesses, Opportunities, Threats). Jenis PLTN ini layak diimplementasikan dalam jangka panjang di Indonesia karena mempunyai aspek keselamatan tinggi, dan mudah dikonstruksi tetapi berbagai kendala masih perlu diselesaikan seperti status teknologi masih desain konsep, SDM belum tersedia, dan belum mendapat sertifikasi dari lembaga yang berwenang.
REVIU在印度尼西亚重新实现THORCON MOLTEN SALT REACTOR。Thorcon Molten Salt Reactor (Thorcon MSR)是Martingle Inc.设计的第四代PLTN类型。美国。这些水冷的液盐柱在高温下工作，接近大气压力。这是一个由4个功率每个250米的模块产生的1000兆兆功率的PLTN。本文的目的是在将在印度尼西亚实施ThorCon MSR时向利益相关者概述ThorCon MSR的可行性。使用的方法是研究Thorcon技术、人力资源、国家参与和干预的各种文献，并通过SWOT分析进行研究。这种类型的PLTN在印尼的长期应用是可行的，因为它具有高度安全、易于建设的结构，但仍需要解决各种障碍，比如技术状态仍然是概念设计的概念，人力资源还没有可用，还没有得到授权机构的认证。
{"title":"Reviu Implementasi Thorcon Molten Salt Reactor di Indonesia","authors":"S. M. Lumbanraja, E. Liun","doi":"10.17146/JPEN.2018.20.1.4083","DOIUrl":"https://doi.org/10.17146/JPEN.2018.20.1.4083","url":null,"abstract":"REVIU IMPLEMENTASI THORCON MOLTEN SALT REACTOR DI INDONESIA. Thorcon Molten Salt Reactor (Thorcon MSR) merupakan jenis PLTN generasi ke-4 yang dirancang oleh Martingle Inc. Amerika Serikat. PLTN ini berbahan bakar, dan berpendingin garam cair , beroperasi pada temperatur tinggi dan tekanannya mendekati tekanan atmosfir. PLTN ini berdaya 1000 MWe yang dihasilkan oleh 4 modul dengan daya masing-masing 250 MWe. Makalah ini bertujuan untuk menguraikan dan memberi gambaran tentang kelayakan ThorCon MSR kepada pemangku kepentingan ketika akan diimplementasikan di Indonesia. Metodologi yang digunakan adalah mempelajari berbagai pustaka tentang teknologi Thorcon, ketersediaan SDM, partisipasi nasional dan regulasinya, serta dilakukan dengan analisis SWOT (Strength, Weaknesses, Opportunities, Threats). Jenis PLTN ini layak diimplementasikan dalam jangka panjang di Indonesia karena mempunyai aspek keselamatan tinggi, dan mudah dikonstruksi tetapi berbagai kendala masih perlu diselesaikan seperti status teknologi masih desain konsep, SDM belum tersedia, dan belum mendapat sertifikasi dari lembaga yang berwenang.","PeriodicalId":179965,"journal":{"name":"Jurnal Pengembangan Energi Nuklir","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2018-09-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"126440963","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2018-06-01DOI: 10.17146/jpen.2018.20.1.4278
S. Dibyo, S. Pinem, V. Wardhani
Conceptual Design On N16 Decay Chamber For Modified TRIGA-2000 With Plate-Type Fuel. the TRIGA-2000 is a research reactor in bandung that will be modified using plate-type fuel. The reactor core cooling system is changed from the natural convection cooling mode to the forced convection mode. The purpose of the study is to assess the conceptual design for the decay chamber of N16 nuclide in the primary cooling system of the reactor. In this design, the hold-up system decays the nuclide of N16 resulted from neutron activation product. In the period of 50 seconds, the activity of N16 (T1/2= 7.13 seconds) decays 7 time from half life to low level. The cube shape of decay chamber is provided a plate with 4 hollows and facility to flush the cavitation bubbles. The decay chamber, which is submerged into the bulk shielding as located outside of the reactor pool. The conceptual design uses the Fluent software compared with the analytical estimation for flow velocity in the decay chamber. The result shows a good agreement range with the analytical estimations. The uniform flow profile can be obtained at the velocity of about 0.4 m/s. Water flow life time of 50 seconds in the decay chamber with the capacity of 3.5 m3 is able to decay the N16 nuclide to low level. This decay chamber is expected to contribute in completing the design of reactor primary coolant system using the forced convection mode.
{"title":"Conceptual Design On N16 Decay Chamber For Modified TRIGA-2000 With Plate-Type Fuel","authors":"S. Dibyo, S. Pinem, V. Wardhani","doi":"10.17146/jpen.2018.20.1.4278","DOIUrl":"https://doi.org/10.17146/jpen.2018.20.1.4278","url":null,"abstract":"Conceptual Design On N16 Decay Chamber For Modified TRIGA-2000 With Plate-Type Fuel. the TRIGA-2000 is a research reactor in bandung that will be modified using plate-type fuel. The reactor core cooling system is changed from the natural convection cooling mode to the forced convection mode. The purpose of the study is to assess the conceptual design for the decay chamber of N16 nuclide in the primary cooling system of the reactor. In this design, the hold-up system decays the nuclide of N16 resulted from neutron activation product. In the period of 50 seconds, the activity of N16 (T1/2= 7.13 seconds) decays 7 time from half life to low level. The cube shape of decay chamber is provided a plate with 4 hollows and facility to flush the cavitation bubbles. The decay chamber, which is submerged into the bulk shielding as located outside of the reactor pool. The conceptual design uses the Fluent software compared with the analytical estimation for flow velocity in the decay chamber. The result shows a good agreement range with the analytical estimations. The uniform flow profile can be obtained at the velocity of about 0.4 m/s. Water flow life time of 50 seconds in the decay chamber with the capacity of 3.5 m3 is able to decay the N16 nuclide to low level. This decay chamber is expected to contribute in completing the design of reactor primary coolant system using the forced convection mode.","PeriodicalId":179965,"journal":{"name":"Jurnal Pengembangan Energi Nuklir","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2018-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124939596","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2018-06-01DOI: 10.17146/jpen.2018.20.1.4305
S. Alimah, J. Mellawati
STUDY OF DISPERSION HAZARD POTENTIAL OF THE LPG STATIONS AROUND THE RDE SITE IN RAINY AND DRY SEASON. There are two LPG station (SPPBE) which are the depot of filling, storage and distribution of Liquid Petroleum Gas (LPG) namely PT. BM and PT. ISR which the distance each are 2,995 and 4,141 km from Experimental Power Reactor (RDE) site with capacity 15 and 30 tons. LPG station is a stationary source, which is one aspect of the external human induced events that need to be analyzed in the preparation of site evaluation reports to obtain site permits. Hazard potential that may occur from the depot LPG are fire, explosion and dispersion of hazardous and toxic gas. The release of LPG due to valve leakage which is then dispersed at a certain dose has potentially harmful to health, even death to the population around the RDE site. The purpose of the study was to know the effect of seasons (rainy and dry) to the potential hazard of LPG dispersion from LPG truck tank valve to the around RDE site. The method of study are collection the atmospheric data such as wind direction and speed, temperature and humidity, collection the station LPG characteristic, such as mass of gas, diameter and length of tank, and valve diameter, etc. The atmospheric data was obtained from Pondok Betung Climatology Station, in dry, transition, and rainy seasons, furthermore data was analyzed using ALOHA software version 5.4.5. The results show dispersion from LPG release due to valve leakage from PT. BM and PT. ISR around the RDE site, in the dry season (April), the transition (January and July) as well as the rainy season (October) does not hazardous to the RDE site. Maximum threat zone occurs in dry season at April (wind speed 1.54 m/s), which reaches radius 179 m with airborne LPG concentration 5500 ppm, radius 111 m with concentration 17000 ppm and radius 71 m with concentration 53000 ppm.
{"title":"Study of Dispersion Hazard Potential of The LPG Stations Around the RDE Site in Rainy and Dry Season","authors":"S. Alimah, J. Mellawati","doi":"10.17146/jpen.2018.20.1.4305","DOIUrl":"https://doi.org/10.17146/jpen.2018.20.1.4305","url":null,"abstract":"STUDY OF DISPERSION HAZARD POTENTIAL OF THE LPG STATIONS AROUND THE RDE SITE IN RAINY AND DRY SEASON. There are two LPG station (SPPBE) which are the depot of filling, storage and distribution of Liquid Petroleum Gas (LPG) namely PT. BM and PT. ISR which the distance each are 2,995 and 4,141 km from Experimental Power Reactor (RDE) site with capacity 15 and 30 tons. LPG station is a stationary source, which is one aspect of the external human induced events that need to be analyzed in the preparation of site evaluation reports to obtain site permits. Hazard potential that may occur from the depot LPG are fire, explosion and dispersion of hazardous and toxic gas. The release of LPG due to valve leakage which is then dispersed at a certain dose has potentially harmful to health, even death to the population around the RDE site. The purpose of the study was to know the effect of seasons (rainy and dry) to the potential hazard of LPG dispersion from LPG truck tank valve to the around RDE site. The method of study are collection the atmospheric data such as wind direction and speed, temperature and humidity, collection the station LPG characteristic, such as mass of gas, diameter and length of tank, and valve diameter, etc. The atmospheric data was obtained from Pondok Betung Climatology Station, in dry, transition, and rainy seasons, furthermore data was analyzed using ALOHA software version 5.4.5. The results show dispersion from LPG release due to valve leakage from PT. BM and PT. ISR around the RDE site, in the dry season (April), the transition (January and July) as well as the rainy season (October) does not hazardous to the RDE site. Maximum threat zone occurs in dry season at April (wind speed 1.54 m/s), which reaches radius 179 m with airborne LPG concentration 5500 ppm, radius 111 m with concentration 17000 ppm and radius 71 m with concentration 53000 ppm.","PeriodicalId":179965,"journal":{"name":"Jurnal Pengembangan Energi Nuklir","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2018-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"125143910","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2018-06-01DOI: 10.17146/jpen.2018.20.1.4257
J. H. Purba, Deswandri Deswandri
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.
{"title":"The Implementation of Importance Measure Approaches for Criticality Analysis in Fault Tree Analysis: A Review","authors":"J. H. Purba, Deswandri Deswandri","doi":"10.17146/jpen.2018.20.1.4257","DOIUrl":"https://doi.org/10.17146/jpen.2018.20.1.4257","url":null,"abstract":"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. ","PeriodicalId":179965,"journal":{"name":"Jurnal Pengembangan Energi Nuklir","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2018-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"116751677","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2018-06-01DOI: 10.17146/jpen.2018.20.1.4095
Mike Susmikanti, Entin Hartini, A. Saepudin, J. Sulistyo
COMPONENT ANALYSIS OF PURIFICATION SYSTEM OF RSG-GAS. Component reliability analysis is required in the aging management of RSG-GAS that has reached an age of 30 years. One of the required analyses is the assessment of the distribution of repair data and the estimation of related parameters. The Primary Purification System (KBE01) and the Purification and Warm Water Layer System (KBE02) are important components of RSG-GAS. By knowing the repair data distribution, the parameters of the most frequently occurring component repair and the average of the repair period can be estimated, so that the required provision of spare parts for the smooth operation of the reactor can be predicted. The purpose of this study is to analyze the components of the KBE01 and KBE02 systems through the data distribution approach using the matching test method. With the matching test, the form of data distribution can be determined, so the parameter of the average component repair period that can be used as a comparison of the maintenance period of the components can be estimated. The repair times of KBE01 and KBE02 in RSG-GAS on Core 52 through Core 88 (2006-2015) were analyzed using goodness-of-fit test. The repair times of AA068 and AP001 KBE01 follow the exponential distribution with average repair times of 631.6 and 451.2 days, respectively. The repair times of WWL and AA002 KBE02 followed an exponential distribution with average repair times of 239.5 days and 888.0 days.
{"title":"Component Analysis of Purification System of RSG-GAS","authors":"Mike Susmikanti, Entin Hartini, A. Saepudin, J. Sulistyo","doi":"10.17146/jpen.2018.20.1.4095","DOIUrl":"https://doi.org/10.17146/jpen.2018.20.1.4095","url":null,"abstract":"COMPONENT ANALYSIS OF PURIFICATION SYSTEM OF RSG-GAS. Component reliability analysis is required in the aging management of RSG-GAS that has reached an age of 30 years. One of the required analyses is the assessment of the distribution of repair data and the estimation of related parameters. The Primary Purification System (KBE01) and the Purification and Warm Water Layer System (KBE02) are important components of RSG-GAS. By knowing the repair data distribution, the parameters of the most frequently occurring component repair and the average of the repair period can be estimated, so that the required provision of spare parts for the smooth operation of the reactor can be predicted. The purpose of this study is to analyze the components of the KBE01 and KBE02 systems through the data distribution approach using the matching test method. With the matching test, the form of data distribution can be determined, so the parameter of the average component repair period that can be used as a comparison of the maintenance period of the components can be estimated. The repair times of KBE01 and KBE02 in RSG-GAS on Core 52 through Core 88 (2006-2015) were analyzed using goodness-of-fit test. The repair times of AA068 and AP001 KBE01 follow the exponential distribution with average repair times of 631.6 and 451.2 days, respectively. The repair times of WWL and AA002 KBE02 followed an exponential distribution with average repair times of 239.5 days and 888.0 days.","PeriodicalId":179965,"journal":{"name":"Jurnal Pengembangan Energi Nuklir","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2018-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124781524","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 1900-01-01DOI: 10.17146/jpen.2021.23.1.6404
Ari Nugroho, E. Kusratmoko, T. L. Indra
PREFERRED SITE SELECTION USING GIS AND AHP: CASE STUDY IN BANGKA ISLAND NPP SITE. Industrial growth affects the increasing demand for electricity in various places, this also occurs on the island of Bangka. So far, electricity supply has only been obtained from fossil fuel power plants with inadequate capacity, unstable flow and depending on fuel supplies from outside the island. For this reason, it is necessary to build a Nuclear Power Plant (PLTN) which is believed to be reliable and able to overcome these problems. In order to prepare a safe and economical nuclear power plant site, influential parameters such as population density, cooling system, land clearing, cut and fill, and granite for the foundation have been analyzed. The novelty of this analysis lies in 2 methods which gradually used before come up with a final decision, namely spatial analysis and pairwise comparison using Geographic Information Systems (GIS) and Analytical Hierarchy Process (AHP), respectively. The scope of study area is based on the site vicinity (1:5.000) scale, located in the districts of West and South Bangka. The siting process refers to the rules set by the International Atomic Energy Agency (IAEA). Based on the final results of the analysis using the expert choice program, the numerical weights for West Bangka and South Bangka were 0.709 and 0.291, respectively, with a consistency value of 0.03.
{"title":"Preferred Site Selection Using GIS and AHP: Case Study in Bangka Island NPP Site","authors":"Ari Nugroho, E. Kusratmoko, T. L. Indra","doi":"10.17146/jpen.2021.23.1.6404","DOIUrl":"https://doi.org/10.17146/jpen.2021.23.1.6404","url":null,"abstract":"PREFERRED SITE SELECTION USING GIS AND AHP: CASE STUDY IN BANGKA ISLAND NPP SITE. Industrial growth affects the increasing demand for electricity in various places, this also occurs on the island of Bangka. So far, electricity supply has only been obtained from fossil fuel power plants with inadequate capacity, unstable flow and depending on fuel supplies from outside the island. For this reason, it is necessary to build a Nuclear Power Plant (PLTN) which is believed to be reliable and able to overcome these problems. In order to prepare a safe and economical nuclear power plant site, influential parameters such as population density, cooling system, land clearing, cut and fill, and granite for the foundation have been analyzed. The novelty of this analysis lies in 2 methods which gradually used before come up with a final decision, namely spatial analysis and pairwise comparison using Geographic Information Systems (GIS) and Analytical Hierarchy Process (AHP), respectively. The scope of study area is based on the site vicinity (1:5.000) scale, located in the districts of West and South Bangka. The siting process refers to the rules set by the International Atomic Energy Agency (IAEA). Based on the final results of the analysis using the expert choice program, the numerical weights for West Bangka and South Bangka were 0.709 and 0.291, respectively, with a consistency value of 0.03.","PeriodicalId":179965,"journal":{"name":"Jurnal Pengembangan Energi Nuklir","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"1900-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"114528329","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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