Symmetry Integrability and Geometry-methods and Applications, 2013
TERMOSIFON. Flowmeter Dynasonics digunakan sebagai alat ukur laju alir pada fasilitas eksperimen ... more TERMOSIFON. Flowmeter Dynasonics digunakan sebagai alat ukur laju alir pada fasilitas eksperimen termosifon dalam rangka pengembangan sistem keselamatan pasif pada PLTN. Untuk dapat digunakan, telah dilakukan optimalisasi pada alat tersebut karena adanya kerusakan pada sistem tampilannya. Optimalisasi dilakukan dengan menggunakan keluaran arus yang diberikan Flowmeter Dynasonic sehingga dapat dibaca oleh sistem akuisisi data National Instruments (DAS-NI). Hal ini dilakukan karena program aplikasi Lab View pada DAS-NI memerlukan persamaan karakterisasi untuk dapat menampilkan data laju alir berdasarkan arus yang diberikan Flowmeter. Pengujian untuk optimalisasi dilakukan dengan membandingkan hasil karakterisasi dan simulasi pada DAS-NI mengunakan kalibrator Jofra. Error pada pembacaan diperoleh dengan membandingkan persamaan karakterisasi dengan simulasi untuk masukan arus yang sama. Dari hasil karakterisasi dan pengujian yang dilakukan diperoleh persamaan karakterisasi y = 177,1x-871,0 dan persamaan pengujian y = 172,8x-851,0 dengan error relatif pembacaan sebesar 3,052297 %. Hasil yang diperoleh menunjukkan bahwa Flowmeter Dynasonics dapat dioptimalkan dengan menggunakan program Lab View.
In the activities of the Pratama Insinas, Ministry of Higher Education technology research in 201... more In the activities of the Pratama Insinas, Ministry of Higher Education technology research in 2018, PTKRN BATAN built a testing facility that simulates a passive cooling system on the reactor core when there is a loss of outside power. The test facility is the Passive-02 System Simulation Facility (FASSIP02).In FASSIP-02 there are several parameters that need to be measured, one of which is temperature. In the measurement of temperature using a K type Thermocouple Connected to the National Instrument 9178 and 9213 modules that use computer programming with LabVIEW software. Temperature measurements need to be characterized.Characterization of type K thermocouples was carried out using thermobaths, 30 type K thermocouples, standard thermocouples,National Instrument modules 9178 and 9213 with computer programming displays using LabVIEW software. The method used for characterization oftype K thermocouples is a fixed temperature comparison method where the results of the temperature con...
Heat pipe is considered to be applied as a passive system to remove residual heat that generated ... more Heat pipe is considered to be applied as a passive system to remove residual heat that generated from reactor core when incident occur or from spent fuel storage pool. The objectives are to characterized the heat transfer phenomena, performance of heat pipe, and as a model for large heat pipe will be applied as passive cooling system on nuclear spent fuel pool storage. In this experimental wickless heat pipe or two-phase closed thermosyphon (TPCT) is used. Variation of heat flux are 611.24 Watt/m<sup>2</sup> - 3291.29 Watt/m<sup>2</sup>. Variation of filling ratio are 45 - 70%. Variation of initial pressure are -62 to -74 cm Hg. Demineralized water is used as working fluid in the TPCT. The results showed that increasing of heat load leads to an increase of evaporation of the working fluid. The optimum filling ratio obtained for 60% of TPCT evaporator volume, and initial pressure variation gave different TPCT wall temperature characteristic. TPCT showed best p...
Pemahaman terhadap manajemen termal apabila terjadi suatu kecelakaan parah reaktor nuklir seperti... more Pemahaman terhadap manajemen termal apabila terjadi suatu kecelakaan parah reaktor nuklir seperti melelehnya bahan bakar dan teras reaktor, menjadi prioritas utama untuk menjaga integritas bejana tekan reaktor. Dengan demikian hasil lelehan bahan bakar dan teras reaktor (debris) tidak keluar dari bejana tekan reaktor dan mengakibatkan dampak lain yang lebih besar ke lingkungan. Salah satu cara yang dilakukan untuk menjaga integritas bejana tekan reaktor adalah dengan melakukan pendinginan terhadap panas berlebih yang dihasilkan akibat dari kecelakaan tersebut. Untuk mempelajari dan mendapatkan pemahaman mengenai hal tersebut, maka dilakukan penelitian mengenai pengaruh suhu awal pelat panas dalam proses quenching (pendinginan secara tiba-tiba) celah sempit rektangular. Penelitian difokuskan pada penentuan suhu rewetting dari pendinginan pelat panas dengan suhu awal pelat 220 0 C, 400 0 C, dan 600 0 C dengan laju aliran air pendingin 0,2 liter/detik. Eksperimen dilakukan dengan mengi...
Telah dilakukkan analisis perubahan alat penukar kalor pada fasilitas Untai UjiBETA. Fasilitas Un... more Telah dilakukkan analisis perubahan alat penukar kalor pada fasilitas Untai UjiBETA. Fasilitas Untai Uji BETA merupakan fasisilitas eksperimen untuk menginvestigasi fenomena thermohidrolikbaik dalam keadaan transien (kecelakaan) ataupun dalam keadaan tunak (operasi normal) sebagaisimulasi sistem pendingin. Adapun komponen dari untai uji BETA terdiri dari pre-heater, pompa primer dansekunder, alat penukar kalor, reservoir tank dan cooling tower. untuk meningkatkan performa alat penukarkalor yang terdapat di UUB adalah dengan cara mengganti alat penukar kalor tersebut. Dengan pergantian alatpenukar kalor maka perlu dilakukan karakterisasi untuk mengetahui performa alat tersebut. Eksperimen ini dilakukan dengan memvariasikan 3 macam debit aliran pada sisi primer, yaitu : 0,377 L/s, 0,472 L/s dan 0,567 L/s adapun untuk debit aliran pada sisi sekunder di beri nilai konstan, yaitu : 1,07 L/s dengan temperatur air60oC . Eksperimen karakterisasi di fokuskan untuk memperoleh hasil efektivita...
Fenomena perpindahan kalor pendidihan pada celah sempit rektangular merupakan fenomena yang berhu... more Fenomena perpindahan kalor pendidihan pada celah sempit rektangular merupakan fenomena yang berhubungan dengan keselamatan reaktor nuklir. Untuk itu perlu dilakukan penelitian tentang hal tersebut di atas agar didapatkan pemahaman yang benar tentang keselamatan reaktor nuklir dari sisi perpindahan kalor pendidihan dan juga dapat berguna bagi perbaikan desain reaktor generasi selanjutnya. Penelitian difokuskan pada perhitungan fluks kalor selama proses pendinginan di celah sempit rektangular berukuran 1,0 mm, dengan suhu awal pelat rektangular 200 o C, 400 o C , dan 600 o C serta laju aliran air pendingin yang masuk ke dalam celah sempit rektangular 0,1 liter/detik, 0,2 liter/detik, dan 0,3 liter/detik. Eksperimen dilakukan dengan menginjeksikan air pada laju aliran tertentu dengan suhu air 85 o C. Data transien suhu hasil pengukuran direkam melalui sistem akuisisi data (DAS, data acquisition system ). Penelitian ini bertujuan untuk memperoleh hubungan fluks kalor kritis dan koefisie...
Untuk meningkatkan keselamatan termal ketika terjadi kecelakaan yang diakibatkan oleh kejadian st... more Untuk meningkatkan keselamatan termal ketika terjadi kecelakaan yang diakibatkan oleh kejadian station blackout, maka dipertimbangkan untuk menggunakan teknologi heat pipe sebagai sistem pendinginan pasif pembuangan panas sisa hasil peluruhan di kolam bahan bakar nuklir bekas. Tujuan penelitian ini adalah untuk mempelajari unjuk kerja termal dari heat pipe yang akan didisain sebagai sistem pendingin pasif di kolam bahan bakar nuklir bekas. Investigasi secara eksperimental dilakukan dengan mempertimbangkan pengaruh tekanan awal di dalam heat pipe dan beban kalor evaporator. Tekanan awal divariasikan pada -54 cm Hg, -64 cm Hg, dan -74 cm Hg. Beban kalor divariasikan pada 1000 W, 1500 W, dan 2000 W. Sedangkan filling ratio, laju aliran massa dan temperature air pendingin diberikan pada nilai konstan 80%, 4 liter/menit, dan 20C. Air demineral digunakan sebagai fluida kerja heat pipe. Air pendingin disirkulasikan pada water jacket sebagai pengambil kalor di condenser. Hasil eksperimen m...
Abstrak—Untai FASSIP-02 terdiri dari beberapa komponen utama yaitu water cooling tank , water hea... more Abstrak—Untai FASSIP-02 terdiri dari beberapa komponen utama yaitu water cooling tank , water heating tank, piping dan tangki ekspansi. FASSIP-02 adalah hasil pengembangan dari FASSIP-01 yang sudah dimiliki terlebih dahulu, FASSIP-01 mempunyai ukuran lebih kecil dibandingkan dengan FASSIP-02 dari segi volume karena volume air pada water heater tank maupun water cooling tank lebih besar, dengan ukuran kolam 3000 x 1000 x 2750 mm. Salah satu aspek yang harus diperhitungkan adalah pembuangan panas sisa hasil peluruhan yang tetap dihasilkan oleh reaktor ketika mengalami kecelakaan. Desain dari komponen-komponen utama tersebut harus dianalisis kemampuannya sebelum dilakukan konstruksi. Studi ini difokuskan pada desain water heating tank sebagai bagian penting dari untai Fasilitas Simulasi Sistem Pasif (FASSIP-02). Tujuan studi ini adalah menghitung kebutuhan pemanas yang akan dipasang dan kekuatan tekanan tangki pada desain water heating tank yang akan dikonstruksi. Metode yang dilakukan...
ABSTRAK ANALISIS KESELAMATAN PELEPASAN PRODUK FISI DARI REAKTOR KARTINI KE LINGKUNGAN. Untuk memp... more ABSTRAK ANALISIS KESELAMATAN PELEPASAN PRODUK FISI DARI REAKTOR KARTINI KE LINGKUNGAN. Untuk mempertinggi tingkat kepercayaan keselamatan pengoperasian reaktor, maka perlu dilakukan analisis keselamatan penyebaran produk fisi ke lingkungan sekitar reaktor saat reaktor beroperasi. Evaluasi keselamatan sistem bantu (auxiliary system) pada sistem ventilasi reaktor Kartini setelah retrofitting telah dilakukan, dengan cara analisis produk fisi dari teras reaktor, melewati sistem ventilasi hingga cerobong, sesuai dengan Design Basis Accident reaktor Kartini. Tujuan penelitian adalah untuk mengetahui besarnya konsentrasi pelepasan paparan radiasi yang keluar dari sistem ventilasi menuju ke lingkungan di sekitar reaktor Kartini dengan mempertimbangkan faktor meteorologi, yaitu suhu dan kecepatan gas buang dari cerobong serta kecepatan angin, digunakan metode Pasquil-Gifford untuk menghitung aktivitas produk fisi yang terlepas ke lingkungan. Hasil pengukuran langsung aktivitas produk fisi da...
Abstrak—Teknologi keselamatan di beberapa desain Pembangkit Listrik Tenaga Nuklir (PLTN) generasi... more Abstrak—Teknologi keselamatan di beberapa desain Pembangkit Listrik Tenaga Nuklir (PLTN) generasi III+ dan IV menggunakan sistem pendingin pasif untuk mengambil panas peluruhan yang dihasilkan oleh reaktor, khususnya pada kondisi kecelakaan. Penelitian dan pengembangan tentang fenomena sistem pendinginan secara pasif telah dilakukan di Pusat Teknologi dan Keselamatan Reaktor Nuklir (PTKRN), Badan Tenaga Nuklir Nasional (BATAN) dengan membuat untai uji Fasilitas Simulasi Sistem Pasif-01 (FASSIP-01). Untai uji tersebut telah dikembangkan dengan mendisain sebuah untai baru yang dinamakan untai uji Fasilitas Simulasi Sistem Pasif-02 (FASSIP-02). Fasilitas uji yang dikembangkan terdiri dari komponen utama yaitu water heating tank (WHT), water cooling tank (WCT), dan sistem pemipaan. Untuk mempelajari fenomena di salah satu komponen utama tersebut, telah dilakukan studi awal dengan simulasi komputer menggunakan software RELAP5. Penelitian yang akan dilakukan adalah melakukan perhitungan k...
GASEOUS RELEASES EVALUATION AND SAFETY PERFORMANCE IMPROVEMENT OF KARTINI RESEARCH REACTOR VENTIL... more GASEOUS RELEASES EVALUATION AND SAFETY PERFORMANCE IMPROVEMENT OF KARTINI RESEARCH REACTOR VENTILATION SYSTEM. The safety performance of Kartini research reactor related to the gaseous releases to the environment has been evaluated. The research covers an evaluation and improvement on the ventilation system and analysis of gas releases dissipating from the reactor building. The method used is calculation of reactor source term and direct measurement of gas release from the reactor stack. The source term analysis showed that the fission product accumulated in the reactor core at the start of operation was 4.838 ´ 106 Ci, after of 5 hours operation it became 3.614 ´ 108 Ci, and after 24 hours decay, the fission product became 4.727 ´ 106 Ci. The N16 activity inside the reactor building is 4.1 ´ 10-10 μCi/cm3 and the Ar41 escaping to the atmosphere is 5.7 ´ 10-12 mCi/cm3, which is lower than limit value for radiation worker of 2 ´ 10-6 μCi/cm3. A sample case by using March 2009 data, t...
View the article online for updates and enhancements. Content from this work may be used under th... more View the article online for updates and enhancements. Content from this work may be used under the terms of the Creative Commons Attribution 3.0 licence. Any further distribution of this work must maintain attribution to the author(s) and the title of the work, journal citation and DOI.
Abstract One of the most important aspects of nuclear safety systems is the reactor cooling syste... more Abstract One of the most important aspects of nuclear safety systems is the reactor cooling system, both in under normal conditions and in accident conditions. Severe accident occurrences due to the failure of the active cooling system, led to the growing importance of the role of passive cooling systems to be implemented in current and future reactor safety systems, in particular. Passive cooling system experiments conducted to understand and determine what parameters affect the natural circulation flow rate, especially local phenomena that occur. So the purpose of this study was conducted to investigate the effects of temperature changes on cold areas and hot area to circulation flow using CFD software and the results of calculations from the experimental data. The experiments were carried out using a facility called the Pre-FASSIP-02 strand, most of which was composed of transparent materials, especially on pipes and heating parts. CFD simulations are also performed to predict the flow rate occurring in natural circulation cases. The analysis was done using water temperature in the near of heater inside water tank which was heated-up during 75 minutes from 26°C in average until 72°C. The result shows that natural circulation flow was increase from -0.035 m/s to 0.218 m/2 using calculation base on experiment data, and using CFD, the flow increased from -0.038 m/2 to 0.327 m/s. The deviation of natural circulation flow between CFD and calculation results are 9.74% for the lower value and 33.17% for the higher value. Mulya Juarsa1, Anhar R. Antariksawan1, Mukhsinun Hadi Kusuma1, Nandy Putra2 and Pryawrata Putera Moniaga3
Natural circulation has an important role in the safety system of a nuclear reactor. It is part o... more Natural circulation has an important role in the safety system of a nuclear reactor. It is part of passive system, which is operated without any external prime mover. The behavior of natural circulation shall be known to be effectively incorporated in the safety system of a nuclear reactor. This paper describes the preliminary results of the study on natural circulation conducted at a vertical rectangular experimental facility FASSIP-01. The RELAP5 code is used to simulate the experiment. The goal is to study the validity of the RELAP 5 model to analyze the natural circulation in FASSIP-01. The simulation could provide a reasonable good results describing the natural circulation established in the rectangular loop of FASSIP-01. However, there are still discrepancies with the experimental results, especially with regard to mass flow rate and temperature distribution in the loop. The model needs be improved and further comprehensive data from experiments are necessitated for better validation, as well.Natural circulation has an important role in the safety system of a nuclear reactor. It is part of passive system, which is operated without any external prime mover. The behavior of natural circulation shall be known to be effectively incorporated in the safety system of a nuclear reactor. This paper describes the preliminary results of the study on natural circulation conducted at a vertical rectangular experimental facility FASSIP-01. The RELAP5 code is used to simulate the experiment. The goal is to study the validity of the RELAP 5 model to analyze the natural circulation in FASSIP-01. The simulation could provide a reasonable good results describing the natural circulation established in the rectangular loop of FASSIP-01. However, there are still discrepancies with the experimental results, especially with regard to mass flow rate and temperature distribution in the loop. The model needs be improved and further comprehensive data from experiments are necessitated for better validation, as well.
The benchmarking of PWR-NPP simulator requires the utilization of 3D code beside 1D code. In case... more The benchmarking of PWR-NPP simulator requires the utilization of 3D code beside 1D code. In case of Emergency Core Cooling System (ECCS) investigations after Loss of Coolant Accident (LOCA), the cooling effectiveness still shows uncertainty between experiment result and 1D thermalhydraulics calculation. It occurs due to the 3-dimmension pressured vessel interior structure is difficult to model in one dimension code. Hence, advanced research on coolant calculation need to be done on the method/technique utilization by applying 3D calculation. One of those methods is Computational Fluid Dynamics (CFD). The purpose of this research is to validate FLUENT 6.3 code for low pressured emergency cooling flow on rectangular narrow channel as a simulation of emergency cooling flow from pressured vessel inlet pipe through narrow channel to downcommer. FLUENT 6.3 validation result shows that the increases of flow-rate and temperature give a trend of the accuracy decrease of the calculation resu...
Boiling heat transfer phenomena on rectangular narrow gap was related to the safety of nuclear re... more Boiling heat transfer phenomena on rectangular narrow gap was related to the safety of nuclear reactors. Research done in order to study the safety of nuclear reactors in particular relating to boiling heat transfer and useful on the improvement of next-generation reactor designs. The research focused on calculation of the heat flux during the cooling process in rectangular narrow gap size 1.0 mm, with initial temperatures 200 °C, 400 °C, and 600 °C, also the flow rates of cooling water 0,1 liters/second, 0,2 liters/second, and 0,3 liters/second. Experiments carried out by injecting water at a certain flow rate with the water temperature 85 ºC. Transient temperature measurement data recorded by the data acquisition system. Transient temperature measurement data is used to calculate the flux of heat gain is then used to obtain the heat transfer coefficient. This research aimed to obtain the correlation between critical heat flux and heat transfer coefficient to changes in temperature...
The research related to thermal management has been significantly inreased, especially for NPP sa... more 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 apparatus constructed 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 70 o C, 80 o C, and 90 o C 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 80 o C, and of about 71.1% and 59.4% at initial temperature of 70 o C. Those results were at higher initial temperature of 90 o C compared with the initial temperature of 90 o C. The average of water mass flow rate increased 81.03% from initial temperatur of 70 o C. 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.
Symmetry Integrability and Geometry-methods and Applications, 2013
TERMOSIFON. Flowmeter Dynasonics digunakan sebagai alat ukur laju alir pada fasilitas eksperimen ... more TERMOSIFON. Flowmeter Dynasonics digunakan sebagai alat ukur laju alir pada fasilitas eksperimen termosifon dalam rangka pengembangan sistem keselamatan pasif pada PLTN. Untuk dapat digunakan, telah dilakukan optimalisasi pada alat tersebut karena adanya kerusakan pada sistem tampilannya. Optimalisasi dilakukan dengan menggunakan keluaran arus yang diberikan Flowmeter Dynasonic sehingga dapat dibaca oleh sistem akuisisi data National Instruments (DAS-NI). Hal ini dilakukan karena program aplikasi Lab View pada DAS-NI memerlukan persamaan karakterisasi untuk dapat menampilkan data laju alir berdasarkan arus yang diberikan Flowmeter. Pengujian untuk optimalisasi dilakukan dengan membandingkan hasil karakterisasi dan simulasi pada DAS-NI mengunakan kalibrator Jofra. Error pada pembacaan diperoleh dengan membandingkan persamaan karakterisasi dengan simulasi untuk masukan arus yang sama. Dari hasil karakterisasi dan pengujian yang dilakukan diperoleh persamaan karakterisasi y = 177,1x-871,0 dan persamaan pengujian y = 172,8x-851,0 dengan error relatif pembacaan sebesar 3,052297 %. Hasil yang diperoleh menunjukkan bahwa Flowmeter Dynasonics dapat dioptimalkan dengan menggunakan program Lab View.
In the activities of the Pratama Insinas, Ministry of Higher Education technology research in 201... more In the activities of the Pratama Insinas, Ministry of Higher Education technology research in 2018, PTKRN BATAN built a testing facility that simulates a passive cooling system on the reactor core when there is a loss of outside power. The test facility is the Passive-02 System Simulation Facility (FASSIP02).In FASSIP-02 there are several parameters that need to be measured, one of which is temperature. In the measurement of temperature using a K type Thermocouple Connected to the National Instrument 9178 and 9213 modules that use computer programming with LabVIEW software. Temperature measurements need to be characterized.Characterization of type K thermocouples was carried out using thermobaths, 30 type K thermocouples, standard thermocouples,National Instrument modules 9178 and 9213 with computer programming displays using LabVIEW software. The method used for characterization oftype K thermocouples is a fixed temperature comparison method where the results of the temperature con...
Heat pipe is considered to be applied as a passive system to remove residual heat that generated ... more Heat pipe is considered to be applied as a passive system to remove residual heat that generated from reactor core when incident occur or from spent fuel storage pool. The objectives are to characterized the heat transfer phenomena, performance of heat pipe, and as a model for large heat pipe will be applied as passive cooling system on nuclear spent fuel pool storage. In this experimental wickless heat pipe or two-phase closed thermosyphon (TPCT) is used. Variation of heat flux are 611.24 Watt/m<sup>2</sup> - 3291.29 Watt/m<sup>2</sup>. Variation of filling ratio are 45 - 70%. Variation of initial pressure are -62 to -74 cm Hg. Demineralized water is used as working fluid in the TPCT. The results showed that increasing of heat load leads to an increase of evaporation of the working fluid. The optimum filling ratio obtained for 60% of TPCT evaporator volume, and initial pressure variation gave different TPCT wall temperature characteristic. TPCT showed best p...
Pemahaman terhadap manajemen termal apabila terjadi suatu kecelakaan parah reaktor nuklir seperti... more Pemahaman terhadap manajemen termal apabila terjadi suatu kecelakaan parah reaktor nuklir seperti melelehnya bahan bakar dan teras reaktor, menjadi prioritas utama untuk menjaga integritas bejana tekan reaktor. Dengan demikian hasil lelehan bahan bakar dan teras reaktor (debris) tidak keluar dari bejana tekan reaktor dan mengakibatkan dampak lain yang lebih besar ke lingkungan. Salah satu cara yang dilakukan untuk menjaga integritas bejana tekan reaktor adalah dengan melakukan pendinginan terhadap panas berlebih yang dihasilkan akibat dari kecelakaan tersebut. Untuk mempelajari dan mendapatkan pemahaman mengenai hal tersebut, maka dilakukan penelitian mengenai pengaruh suhu awal pelat panas dalam proses quenching (pendinginan secara tiba-tiba) celah sempit rektangular. Penelitian difokuskan pada penentuan suhu rewetting dari pendinginan pelat panas dengan suhu awal pelat 220 0 C, 400 0 C, dan 600 0 C dengan laju aliran air pendingin 0,2 liter/detik. Eksperimen dilakukan dengan mengi...
Telah dilakukkan analisis perubahan alat penukar kalor pada fasilitas Untai UjiBETA. Fasilitas Un... more Telah dilakukkan analisis perubahan alat penukar kalor pada fasilitas Untai UjiBETA. Fasilitas Untai Uji BETA merupakan fasisilitas eksperimen untuk menginvestigasi fenomena thermohidrolikbaik dalam keadaan transien (kecelakaan) ataupun dalam keadaan tunak (operasi normal) sebagaisimulasi sistem pendingin. Adapun komponen dari untai uji BETA terdiri dari pre-heater, pompa primer dansekunder, alat penukar kalor, reservoir tank dan cooling tower. untuk meningkatkan performa alat penukarkalor yang terdapat di UUB adalah dengan cara mengganti alat penukar kalor tersebut. Dengan pergantian alatpenukar kalor maka perlu dilakukan karakterisasi untuk mengetahui performa alat tersebut. Eksperimen ini dilakukan dengan memvariasikan 3 macam debit aliran pada sisi primer, yaitu : 0,377 L/s, 0,472 L/s dan 0,567 L/s adapun untuk debit aliran pada sisi sekunder di beri nilai konstan, yaitu : 1,07 L/s dengan temperatur air60oC . Eksperimen karakterisasi di fokuskan untuk memperoleh hasil efektivita...
Fenomena perpindahan kalor pendidihan pada celah sempit rektangular merupakan fenomena yang berhu... more Fenomena perpindahan kalor pendidihan pada celah sempit rektangular merupakan fenomena yang berhubungan dengan keselamatan reaktor nuklir. Untuk itu perlu dilakukan penelitian tentang hal tersebut di atas agar didapatkan pemahaman yang benar tentang keselamatan reaktor nuklir dari sisi perpindahan kalor pendidihan dan juga dapat berguna bagi perbaikan desain reaktor generasi selanjutnya. Penelitian difokuskan pada perhitungan fluks kalor selama proses pendinginan di celah sempit rektangular berukuran 1,0 mm, dengan suhu awal pelat rektangular 200 o C, 400 o C , dan 600 o C serta laju aliran air pendingin yang masuk ke dalam celah sempit rektangular 0,1 liter/detik, 0,2 liter/detik, dan 0,3 liter/detik. Eksperimen dilakukan dengan menginjeksikan air pada laju aliran tertentu dengan suhu air 85 o C. Data transien suhu hasil pengukuran direkam melalui sistem akuisisi data (DAS, data acquisition system ). Penelitian ini bertujuan untuk memperoleh hubungan fluks kalor kritis dan koefisie...
Untuk meningkatkan keselamatan termal ketika terjadi kecelakaan yang diakibatkan oleh kejadian st... more Untuk meningkatkan keselamatan termal ketika terjadi kecelakaan yang diakibatkan oleh kejadian station blackout, maka dipertimbangkan untuk menggunakan teknologi heat pipe sebagai sistem pendinginan pasif pembuangan panas sisa hasil peluruhan di kolam bahan bakar nuklir bekas. Tujuan penelitian ini adalah untuk mempelajari unjuk kerja termal dari heat pipe yang akan didisain sebagai sistem pendingin pasif di kolam bahan bakar nuklir bekas. Investigasi secara eksperimental dilakukan dengan mempertimbangkan pengaruh tekanan awal di dalam heat pipe dan beban kalor evaporator. Tekanan awal divariasikan pada -54 cm Hg, -64 cm Hg, dan -74 cm Hg. Beban kalor divariasikan pada 1000 W, 1500 W, dan 2000 W. Sedangkan filling ratio, laju aliran massa dan temperature air pendingin diberikan pada nilai konstan 80%, 4 liter/menit, dan 20C. Air demineral digunakan sebagai fluida kerja heat pipe. Air pendingin disirkulasikan pada water jacket sebagai pengambil kalor di condenser. Hasil eksperimen m...
Abstrak—Untai FASSIP-02 terdiri dari beberapa komponen utama yaitu water cooling tank , water hea... more Abstrak—Untai FASSIP-02 terdiri dari beberapa komponen utama yaitu water cooling tank , water heating tank, piping dan tangki ekspansi. FASSIP-02 adalah hasil pengembangan dari FASSIP-01 yang sudah dimiliki terlebih dahulu, FASSIP-01 mempunyai ukuran lebih kecil dibandingkan dengan FASSIP-02 dari segi volume karena volume air pada water heater tank maupun water cooling tank lebih besar, dengan ukuran kolam 3000 x 1000 x 2750 mm. Salah satu aspek yang harus diperhitungkan adalah pembuangan panas sisa hasil peluruhan yang tetap dihasilkan oleh reaktor ketika mengalami kecelakaan. Desain dari komponen-komponen utama tersebut harus dianalisis kemampuannya sebelum dilakukan konstruksi. Studi ini difokuskan pada desain water heating tank sebagai bagian penting dari untai Fasilitas Simulasi Sistem Pasif (FASSIP-02). Tujuan studi ini adalah menghitung kebutuhan pemanas yang akan dipasang dan kekuatan tekanan tangki pada desain water heating tank yang akan dikonstruksi. Metode yang dilakukan...
ABSTRAK ANALISIS KESELAMATAN PELEPASAN PRODUK FISI DARI REAKTOR KARTINI KE LINGKUNGAN. Untuk memp... more ABSTRAK ANALISIS KESELAMATAN PELEPASAN PRODUK FISI DARI REAKTOR KARTINI KE LINGKUNGAN. Untuk mempertinggi tingkat kepercayaan keselamatan pengoperasian reaktor, maka perlu dilakukan analisis keselamatan penyebaran produk fisi ke lingkungan sekitar reaktor saat reaktor beroperasi. Evaluasi keselamatan sistem bantu (auxiliary system) pada sistem ventilasi reaktor Kartini setelah retrofitting telah dilakukan, dengan cara analisis produk fisi dari teras reaktor, melewati sistem ventilasi hingga cerobong, sesuai dengan Design Basis Accident reaktor Kartini. Tujuan penelitian adalah untuk mengetahui besarnya konsentrasi pelepasan paparan radiasi yang keluar dari sistem ventilasi menuju ke lingkungan di sekitar reaktor Kartini dengan mempertimbangkan faktor meteorologi, yaitu suhu dan kecepatan gas buang dari cerobong serta kecepatan angin, digunakan metode Pasquil-Gifford untuk menghitung aktivitas produk fisi yang terlepas ke lingkungan. Hasil pengukuran langsung aktivitas produk fisi da...
Abstrak—Teknologi keselamatan di beberapa desain Pembangkit Listrik Tenaga Nuklir (PLTN) generasi... more Abstrak—Teknologi keselamatan di beberapa desain Pembangkit Listrik Tenaga Nuklir (PLTN) generasi III+ dan IV menggunakan sistem pendingin pasif untuk mengambil panas peluruhan yang dihasilkan oleh reaktor, khususnya pada kondisi kecelakaan. Penelitian dan pengembangan tentang fenomena sistem pendinginan secara pasif telah dilakukan di Pusat Teknologi dan Keselamatan Reaktor Nuklir (PTKRN), Badan Tenaga Nuklir Nasional (BATAN) dengan membuat untai uji Fasilitas Simulasi Sistem Pasif-01 (FASSIP-01). Untai uji tersebut telah dikembangkan dengan mendisain sebuah untai baru yang dinamakan untai uji Fasilitas Simulasi Sistem Pasif-02 (FASSIP-02). Fasilitas uji yang dikembangkan terdiri dari komponen utama yaitu water heating tank (WHT), water cooling tank (WCT), dan sistem pemipaan. Untuk mempelajari fenomena di salah satu komponen utama tersebut, telah dilakukan studi awal dengan simulasi komputer menggunakan software RELAP5. Penelitian yang akan dilakukan adalah melakukan perhitungan k...
GASEOUS RELEASES EVALUATION AND SAFETY PERFORMANCE IMPROVEMENT OF KARTINI RESEARCH REACTOR VENTIL... more GASEOUS RELEASES EVALUATION AND SAFETY PERFORMANCE IMPROVEMENT OF KARTINI RESEARCH REACTOR VENTILATION SYSTEM. The safety performance of Kartini research reactor related to the gaseous releases to the environment has been evaluated. The research covers an evaluation and improvement on the ventilation system and analysis of gas releases dissipating from the reactor building. The method used is calculation of reactor source term and direct measurement of gas release from the reactor stack. The source term analysis showed that the fission product accumulated in the reactor core at the start of operation was 4.838 ´ 106 Ci, after of 5 hours operation it became 3.614 ´ 108 Ci, and after 24 hours decay, the fission product became 4.727 ´ 106 Ci. The N16 activity inside the reactor building is 4.1 ´ 10-10 μCi/cm3 and the Ar41 escaping to the atmosphere is 5.7 ´ 10-12 mCi/cm3, which is lower than limit value for radiation worker of 2 ´ 10-6 μCi/cm3. A sample case by using March 2009 data, t...
View the article online for updates and enhancements. Content from this work may be used under th... more View the article online for updates and enhancements. Content from this work may be used under the terms of the Creative Commons Attribution 3.0 licence. Any further distribution of this work must maintain attribution to the author(s) and the title of the work, journal citation and DOI.
Abstract One of the most important aspects of nuclear safety systems is the reactor cooling syste... more Abstract One of the most important aspects of nuclear safety systems is the reactor cooling system, both in under normal conditions and in accident conditions. Severe accident occurrences due to the failure of the active cooling system, led to the growing importance of the role of passive cooling systems to be implemented in current and future reactor safety systems, in particular. Passive cooling system experiments conducted to understand and determine what parameters affect the natural circulation flow rate, especially local phenomena that occur. So the purpose of this study was conducted to investigate the effects of temperature changes on cold areas and hot area to circulation flow using CFD software and the results of calculations from the experimental data. The experiments were carried out using a facility called the Pre-FASSIP-02 strand, most of which was composed of transparent materials, especially on pipes and heating parts. CFD simulations are also performed to predict the flow rate occurring in natural circulation cases. The analysis was done using water temperature in the near of heater inside water tank which was heated-up during 75 minutes from 26°C in average until 72°C. The result shows that natural circulation flow was increase from -0.035 m/s to 0.218 m/2 using calculation base on experiment data, and using CFD, the flow increased from -0.038 m/2 to 0.327 m/s. The deviation of natural circulation flow between CFD and calculation results are 9.74% for the lower value and 33.17% for the higher value. Mulya Juarsa1, Anhar R. Antariksawan1, Mukhsinun Hadi Kusuma1, Nandy Putra2 and Pryawrata Putera Moniaga3
Natural circulation has an important role in the safety system of a nuclear reactor. It is part o... more Natural circulation has an important role in the safety system of a nuclear reactor. It is part of passive system, which is operated without any external prime mover. The behavior of natural circulation shall be known to be effectively incorporated in the safety system of a nuclear reactor. This paper describes the preliminary results of the study on natural circulation conducted at a vertical rectangular experimental facility FASSIP-01. The RELAP5 code is used to simulate the experiment. The goal is to study the validity of the RELAP 5 model to analyze the natural circulation in FASSIP-01. The simulation could provide a reasonable good results describing the natural circulation established in the rectangular loop of FASSIP-01. However, there are still discrepancies with the experimental results, especially with regard to mass flow rate and temperature distribution in the loop. The model needs be improved and further comprehensive data from experiments are necessitated for better validation, as well.Natural circulation has an important role in the safety system of a nuclear reactor. It is part of passive system, which is operated without any external prime mover. The behavior of natural circulation shall be known to be effectively incorporated in the safety system of a nuclear reactor. This paper describes the preliminary results of the study on natural circulation conducted at a vertical rectangular experimental facility FASSIP-01. The RELAP5 code is used to simulate the experiment. The goal is to study the validity of the RELAP 5 model to analyze the natural circulation in FASSIP-01. The simulation could provide a reasonable good results describing the natural circulation established in the rectangular loop of FASSIP-01. However, there are still discrepancies with the experimental results, especially with regard to mass flow rate and temperature distribution in the loop. The model needs be improved and further comprehensive data from experiments are necessitated for better validation, as well.
The benchmarking of PWR-NPP simulator requires the utilization of 3D code beside 1D code. In case... more The benchmarking of PWR-NPP simulator requires the utilization of 3D code beside 1D code. In case of Emergency Core Cooling System (ECCS) investigations after Loss of Coolant Accident (LOCA), the cooling effectiveness still shows uncertainty between experiment result and 1D thermalhydraulics calculation. It occurs due to the 3-dimmension pressured vessel interior structure is difficult to model in one dimension code. Hence, advanced research on coolant calculation need to be done on the method/technique utilization by applying 3D calculation. One of those methods is Computational Fluid Dynamics (CFD). The purpose of this research is to validate FLUENT 6.3 code for low pressured emergency cooling flow on rectangular narrow channel as a simulation of emergency cooling flow from pressured vessel inlet pipe through narrow channel to downcommer. FLUENT 6.3 validation result shows that the increases of flow-rate and temperature give a trend of the accuracy decrease of the calculation resu...
Boiling heat transfer phenomena on rectangular narrow gap was related to the safety of nuclear re... more Boiling heat transfer phenomena on rectangular narrow gap was related to the safety of nuclear reactors. Research done in order to study the safety of nuclear reactors in particular relating to boiling heat transfer and useful on the improvement of next-generation reactor designs. The research focused on calculation of the heat flux during the cooling process in rectangular narrow gap size 1.0 mm, with initial temperatures 200 °C, 400 °C, and 600 °C, also the flow rates of cooling water 0,1 liters/second, 0,2 liters/second, and 0,3 liters/second. Experiments carried out by injecting water at a certain flow rate with the water temperature 85 ºC. Transient temperature measurement data recorded by the data acquisition system. Transient temperature measurement data is used to calculate the flux of heat gain is then used to obtain the heat transfer coefficient. This research aimed to obtain the correlation between critical heat flux and heat transfer coefficient to changes in temperature...
The research related to thermal management has been significantly inreased, especially for NPP sa... more 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 apparatus constructed 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 70 o C, 80 o C, and 90 o C 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 80 o C, and of about 71.1% and 59.4% at initial temperature of 70 o C. Those results were at higher initial temperature of 90 o C compared with the initial temperature of 90 o C. The average of water mass flow rate increased 81.03% from initial temperatur of 70 o C. 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.
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Papers by M. Hadi Kusuma