Articles

Found 20 Documents
Search

INVESTIGATION OF RDE THERMAL PARAMETERS CHANGES IN RESPONSE TO LONG-TERM STATION BLACK OUT Tjahjono, Hendro
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 19, No 2 (2017): Juni 2017
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (3363.532 KB) | DOI: 10.17146/tdm.2017.19.2.3258

Abstract

Due to long-term station black out (SBO) of the RDE (Experimental Power Reactor), the residual heat from the core will be removed to a residual heat removal system (RHRS). The objective of this study is to know the transient characteristic of RDE thermal parameters in response to the loss of residual heat removing ability for long-term. To achieve this objective, an analysis model of reactor thermal parameters changes during SBO, using Matlab program to simulate heat transfer equations of conduction, convection and radiation has been performed. Using this program, the changes of RDE thermal parameters until 800 hours after reactor trip have been analyzed. It is concluded that, in long-term SBO condition, the reactor is still safe with the maximum core temperature of 1140°C, which is still far under the safety limit of 1600°C as stated in the design criteria. More attentions are needed to be taken with the increasing of concrete temperature up to 600°C when the water storage is empty. Therefore, the availability of water in the RHRS shall absolutely be maintained.Keywords: experimental power reactor, residual heat removal, transient, Matlab. INVESTIGASI PERUBAHAN PARAMETER TERMAL RDE PADA KONDISI KEHILANGAN CATU DAYA LISTRIK DALAM JANGKA PANJANG. Akibat kehilangan catu daya listrik luar pada Reaktor Daya Eksperimental (RDE), panas sisa dari reaktor dibuang ke suatu sistem pembuang panas sisa. Penelitian ini bertujuan untuk mengetahui karakteristik transien parameter termal RDE ketika terjadi kegagalan pembuangan kalor sisa tersebut dalam jangka panjang. Untuk mencapai tujuan tersebut telah disusun model analisis perubahan parameter termal reaktor ketika terjadi Station Black Out (SBO) menggunakan pemrograman Matlab dengan melibatkan persamaan-persamaan perpindahan kalor secara konduksi, konveksi dan radiasi. Dengan menggunakan program ini perubahan parameter termal RDE hingga 800 jam setelah reaktor trip telah dianalisis. Disimpulkan bahwa pada kondisi SBO dalam jangka panjang tersebut, reaktor masih tetap aman dengan temperatur maksimum teras sebesar 1140 °C, yaitu masih jauh di bawah batas aman 1600 °C yang telah ditetapkan dalam kriteria desain. Perlu diperhatikan adanya peningkatan temperatur beton hingga 600 °C jika air pendingin sudah habis. Oleh karena itu, ketersediaan air pendingin di sistem pembuang panas sisa mutlak harus dijaga.Kata kunci: reaktor daya eksperimental, pembuang panas sisa, transien, Matlab.
SISTEM PENGENDALI ARUS START MOTOR INDUKSI PHASA TIGA DENGAN VARIASI BEBAN Yusnita, Yusnita; Tjahjono, Hendro
Jurnal Teknik Elektro - ITP Vol 1, No 2 (2012): Jurnal Teknik Elektro
Publisher : ITP Press

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (96.698 KB)

Abstract

Penelitian ini dimaksudkan untuk memberikan suata cara sederhana dalam mengendalikan arus start motor induksi 3-fasa. Penelitian ini dibantu oleh program komputer Matlab untuk menganalisa kinerja motor selama proses start dan operasi. Motor yang digunakan pada penelitian ini adalah motoro induksi 3-fasa, 1500 W, 380 V, 4 kutup, 50 Hz and 1400 rpm. Dari hasil penelitian yang telah dilakukan menunjukkan bahwa untuk mengendalikan arus start sebaiknya pada awal start motor diberi tegangan 26.32% dari tegangan nominalnya,kemudian dinaikan secara bertahap menjadi 69% and 100% dari tegangan nominalnya.
EFFECT OF AIR CONDITION ON AP-1000 CONTAINMENT COOLING PERFORMANCE IN STATION BLACK OUT ACCIDENT Tjahjono, Hendro
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 17, No 3 (2015): Oktober 2015
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (571.919 KB) | DOI: 10.17146/tdm.2015.17.3.2323

Abstract

ABSTRACT EFFECT OF AIR CONDITION ON AP-1000 CONTAINMENT COOLING PERFORMANCE IN STATION BLACK OUT ACCIDENT. AP1000 reactor is a nuclear power plant generation III+ 1000 MWe which apply passive cooling concept to anticipate accidents triggered by the extinction of the entire supply of electrical power or Station Black Out (SBO). In the AP1000 reactor, decay heat disposal mechanism conducted passively through the PRHR-IRWST and subsequently forwarded to the reactor containment. Containment externally cooled through natural convection in the air gap and through evaporation cooling water poured on the outer surface of the containment wall. The mechanism of evaporation of water into the air outside is strongly influenced by the conditions of humidity and air temperature. The purpose of this study was to determine the extent of the influence of the air condition on cooling capabilities of the AP1000 containment. The method used is to perform simulations using Matlab-based analytical calculation model capable of estimating the power of heat transfered. The simulation results showed a decrease in power up to  5% for relative humidity rose from 10% to 95%, while the variation of air temperature of 10 °C to 40°C, the power will decrease up to 15%. It can be concluded that the effect of air temperature increase is much more significant in lowering the containment cooling ability compared with the increase of humidity. Keywords: containment cooling, AP1000, air condition, SBO   ABSTRAK PENGARUH KONDISI UDARA TERHADAP KINERJA PENDINGINAN SUNGKUP AP-1000 DALAM KECELAKAAN STATION BLACK OUT. Reaktor AP-1000 merupakan PLTN generasi III+ berdaya 1000 MWe yang menerapkan konsep pendinginan pasif untuk mengantisipasi terjadinya kecelakaan yang dipicu oleh padamnya seluruh suplai daya listrik atau dikenal dengan Station Black Out (SBO). Pada reaktor AP-1000, mekanisme pembuangan kalor peluruhan dilakukan secara pasif melalui PRHR yang diteruskan ke IRWST dan selanjutnya pada sungkup reaktor. Sungkup didinginkan secara eksternal melalui konveksi alamiah pada celah udara dan melalui penguapan air pendingin yang diguyurkan di permukaan luar dinding sungkup. Mekanisme penguapan air ke udara luar sangat dipengaruhi oleh kondisi kelembaban dan temperatur udara. Tujuan dari penelitian ini adalah untuk mengetahui sejauh mana pengaruh kondisi udara tersebut terhadap kemampuan pendinginan dari sungkup AP1000. Metode yang digunakan adalah dengan melakukan simulasi menggunakan model perhitungan analitis berbasis Matlab yang mampu mengestimasi daya kalor yang dievakuasi. Hasil simulasi menunjukkan adanya penurunan daya hingga 5% untuk kelembaban relatif naik dari 10% hingga 95%, sedangkan untuk variasi temperatur udara dari 10°C hingga 40°C, daya akan menurun hingga 15%.  Dapat disimpulkan bahwa pengaruh kenaikan temperatur udara jauh lebih signifikan dalam menurunkan kemampuan pendinginan sungkup dibandingkan dengan naiknya kelembaban. Kata kunci: pendinginan sungkup, AP1000,  kondisi udara, SBO
REACTOR CAVITY COOLING SYSTEM WITH PASSIVE SAFETY FEATURES ON RDE: THERMAL ANALYSIS DURING ACCIDENT Kusumastuti, Rahayu; Sriyono, Sriyono; Juarsa, Mulya; Tjahjono, Hendro; Irianto, I. D.; Setiadipura, Topan; Salimy, D. H.; Hafid, A.
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 21, No 2 (2019): JUNI 2019
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (1878.724 KB) | DOI: 10.17146/tdm.2019.21.2.5499

Abstract

Reaktor Daya Eksperimental (RDE) is an experimental power reactor based on HTGR technology that implements inherent safety system. Its safety systems are in compliance with ?defense in depth? philosophy. RDE is also equipped with reactor cavity cooling system (RCCS) used to remove the heat transferred from the reactor vessel to the containment structure. The RCCS is designed to fulfil this role by maintain the reactor vessel under the maximum allowable temperature during normal operation and protecting the containment structure in the event of failure of all passive cooling systems. The performance and reliability of the RCCS, therefore, are considered as critical factors in determining maximum design power level related to heat removal. RCCS for RDE will use a novel shape to efficiently remove the heat released from the RPV through thermal radiation and natural convection. This paper discusses the calculation of RCCS thermal analysis during accident. The RPV temperature must be maintained below 65ºC. The accident is assumed that there is no electricity from diesel generator supplied to the blower. The methodology used is based on the calculation of mathematical model of the RCCS in the passive mode. The heat is released through cavity by natural convection, in which the RCCS is capable to withdraw the heat at the rate of 50.54 kW per hour.Keywords: Passive safety, RCCS, RDE, Thermal analysis
INVESTIGASI KARAKTERISTIK TERMOHIDROLIKA TERAS REAKTOR DAYA KECIL DENGAN PENDINGINAN SIRKULASI ALAM MENGGUNAKAN RELAP5 Susyadi, Susyadi; Tjahjono, Hendro; Dibyo, Sukmanto; Pane, Jupiter Sitorus
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 18, No 1 (2016): Februari 2016
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (760.867 KB) | DOI: 10.17146/tdm.2016.18.1.2330

Abstract

ABSTRAK INVESTIGASI KARAKTERISTIK TERMOHIDROLIKA TERAS REAKTOR DAYA KECIL DENGAN PENDINGINAN SIRKULASI ALAM MENGGUNAKAN RELAP5. Reaktor modular daya-kecil (small modular reactor, SMR) memiliki prospek tinggi untuk dibangun di Indonesia. Keluaran dayanya yang relatif kecil dan disainnya yang kompak serta dapat dikonstruksi secara modular memberikan keunggulan fleksibilitas pembangunan yang lebih baik dibanding reaktor konvensional berdaya besar. Disain sistem reaktor kategori ini sangat bervariasi, salah satu diantaranya adalah jenis reaktor air tekan (pressurized water reactor, PWR) yang menerapkan sirkulasi alamiah pada sistem pendingin primernya. Selain itu reaktor ini juga memiliki teras (core) lebih pendek dibanding PWR konvensional. Dari kedua perbedaan tersebut maka terdapat kemungkinan perbedaan pola perpindahan panas yang dapat berimplikasi terhadap keselamatan secara keseluruhan. Oleh karena itu, pada penelitian ini dilakukan investigasi terhadap karakteristik termohidrolika teras reaktor tersebut khususnya karakteristik temperatur fluida dan bahan bakar serta laju alir fluidanya. Tujuannya adalah untuk mengetahui perbedaan marjin keselamatan temperatur teras reaktor bila dibanding dengan PWR konvensional. Investigasi dilakukan dengan menggunakan program RELAP5, dimana secara parsial teras reaktor dimodelkan menggunakan model-model generik yang ada pada program dan dilakukan beberapa perhitungan kondisi tunak. Hasil perhitungan menunjukkan bahwa saat beroperasi pada daya nominalnya, reaktor modular ini memiliki margin temperatur pendidihan sebesar 2K lebih baik dibanding reaktor konvensional. Selain itu, keunggulan marjin keselamatan reaktor modular daya-kecil ini juga ditunjukkan dari naiknya laju alir mengikuti kenaikan dayanya yang berarti memiliki sifat keselamatan yang melekat (inherent safety). Kata kunci: reaktor modular daya-kecil, PWR, sirkulasi alam, RELAP5, termohidrolika   ABSTRACT INVESTIGATION ON CORE THERMAL HYDRAULIC CHARACTERISTICS OF SMALL MODULAR REACTOR WITH NATURAL CIRCULATION COOLING USING RELAP5. Small modular reactor (SMR ) is very prospective to be deployed in Indonesia. Its low output power, compact design and capability to be constructed modularly provide better deployment flexibility compared to a large conventional reactor. There are various designs of SMRs, one of them implements natural circulation for its primary cooling system or in other words the reactor uses no primary pumps. Besides, the dimension of fuel element is shorter than the one used by large reactor. These two aspects may produce different heat transfer behavior, which could lead to a safety implication.  For that reason, this research investigates thermal hydraulic characteristics of the core of SMR with naturally circulating coolant, especially on the fuel and coolant temperatures and mass flow rate. The purpose is to identify the thermal safety margin difference of the reactor compared with conventional PWR.  The investigation was performed using RELAP5 in which the core was partially represented by means of generic models of the program and continued with steady state calculations. The result shows that during nominal power operation, the reactor has better of 2K  degree for boiling temperature margin than the large conventional PWR. In addition, the excellence of SMR safety margin was shown by the increase of primary coolant flow rate following the increase of power, which means that the reactor has a distinctive inherent safety. Keywords: small modular reaktor, PWR, natural circulation, RELAP5, thermal-hydraulic
OPTIMASI PENDINGINAN EKSTERNAL PADA MODEL SUNGKUP PWR-1000 MENGGUNAKAN METODE ESTIMASI ANALITIK Tjahjono, Hendro
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 16, No 2 (2014): Juni 2014
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (396.1 KB)

Abstract

Sungkup reaktor merupakan benteng terakhir dalam menahan terlepasnya zat-zat radioaktif ke lingkungan ketika terjadi suatu kecelakaan reaktor. Oleh karena itu integritasnya harus selalu dipertahankan yang antara lain dilakukan dengan cara mencegah dilampauinya batas desain tekanan dan temperatur yang bisa terjadi pada kondisi kecelakaan melalui pendinginan sungkup yang mencukupi. Pada reaktor generasi III+ yang menerapkan konsep pendinginan pasif seperti AP1000, sungkup didinginkan secara eksternal melalui konveksi alamiah pada celah udara dan guyuran air pendingin di permukaan luar sungkup. Karakteristik pendinginan eksternal ini akan diteliti secara eksperimental melalui model sungkup PWR1000 berskala 1/40. Tujuan dari penelitian ini adalah untuk mengetahui nilai debit optimal yang diperlukan dalam pendinginan model sungkup sebelum konfirmasi secara eksperimental dilakukan. Metode yang digunakan adalah dengan melakukan pemodelan analitis dan pemrograman berbasis Matlab yang mampu mengestimasi nilai-nilai parameter pendinginan eksternal seperti laju alir, temperatur dan daya kalor yang dievakuasi. Penerapan program pada sungkup AP1000 juga dilakukan untuk bisa dibandingkan dengan data desain. Hasilnya menunjukkan kesesuaian dengan data desain sungkup AP1000 dengan debit optimal sebesar 9,5 liter/detik yang mampu mengevakuasi kalor sebesar 21,6 MW. Sedangkan pada model sungkup diperoleh debit optimal sebesar 22 cc/detik yang mampu mengevakuasi kalor sebesar 37 KW. Disimpulkan bahwa dengan penelitian ini karakteristik pendinginan eksternal sungkup reaktor PWR mampu diestimasi dan bersamaan dengan itu dapat diketahui nilai optimal dari debit pendingin yang diperlukan.Kata kunci: pendinginan eksternal, sungkup PWR, estimasi analitik, AP1000.    Reactor containment is the last barrier in avoiding the release of radioactive substances into the environment in the event of a reactor accident. Therefore, its integrity must always be maintained, among others, performed in a manner to prevent the exceeding of pressure and temperature design limit that could occur in an accident, through adequate containment cooling. In generation III + reactors which passive cooling concepts are applied such as the AP1000, the containment is externally cooled by natural convection in the air gap and a splash of cooling water in the outer surface. External cooling characteristics will be investigated experimentally through PWR1000 containment models of 1/40 scale. The purpose of this research is to determine the optimal flow of cooling required in the model before confirming experimentally performed. The method used is to perform analytical modeling and programming based on Matlab which is able to estimate the values of external cooling parameters such as flow rate, temperature and heat power evacuated. Implementation of the program on the AP1000 containment is also performed to be compared with the design data. The results shows the conformity with the AP1000 containment design data with optimal flow of 9.5 liters/sec that is able to evacuate the heat of 21.6 MW. While for the containment model, the optimal flow obtained at 22 cc / sec which is capable of evacuating the heat by 37 KW. This study concluded that the characteristics of the external cooling of PWR containment could be estimated and in conjunction, the optimal cooling flow required can be determined. Keywords: external cooling, PWR containment, analytical estimation, AP1000.
DEVELOPMENT OF EXPERIMENTAL POWER REACTOR (EPR) MODEL FOR SAFETY ANALYSES USING RELAP5 Ekariansyah, Andi Sofrany; Subekti, Muhammad; Widodo, Surip; Tjahjono, Hendro; Susyadi, Susyadi; Wahyono, Puradwi Ismu; Budianto, Anwar
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 21, No 2 (2019): JUNI 2019
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (1853.799 KB) | DOI: 10.17146/tdm.2019.21.2.5449

Abstract

Pebble bed reactor design, classified as the high temperature gas-cooled reactor (HTGR), is currently being part of BATAN main program to promote nuclear energy by starting the Experimental Power Reactor (EPR) program since 2015. Starting from 2018, the detail design document has to be submitted into nuclear regulatory body for further assessment. Therefore results of design analysis have to be supplemented by performing a design evaluation, which can be achieved by developing the model of the EPR.  The development is performed using RELAP5/SCDAP/Mod.3.4 as the thermal-hydraulic analysis code validated for the light-water reactor having module for the pebble fuel element and non-condensable helium gas. Methodology of model development consists of defining the helium flow path inside the reactor pressure vessel, modelling of pebble bed core including its power distribution, and modelling of reflector components to be simulated under 100 % core power. The developed EPR model results in design parameters, which confirm the main thermal data of the EPR, including the pebble and reflector temperatures. The peak pebble temperature is calculated to be 1,375 °C, which requires further investigations in the model accuracy, since the reference values are around 1,015 °C, even it is below the pebble temperature limit. For safety analysis, the EPR model can be used under nominal core flow condition, which produces more conservative results by paying attention on the RELAP5 specific modules for the pebble bed-gas cooled system.Keywords: experimental power reactor, development, RELAP5, steady-state
PENGEMBANGAN MODEL UNTUK SIMULASI KESELAMATAN REAKTOR PWR 1000 MWE GENERASI III+ MENGGUNAKAN PROGRAM KOMPUTER RELAP5 Ekariansyah, Andi Sofrany; Widodo, Surip; Susyadi, Susyadi; Sony Tjahyani, D.T.; Tjahjono, Hendro
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 13, No 1 (2011): Pebruari 2011
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (741.617 KB)

Abstract

Reaktor daya PWR AP1000 yang didesain oleh Westinghouse adalah reaktor Generasi III+ pertama yang telah menerima persetujuan desain dari U.S. Nuclear Regulatory Commission (NRC). Saat ini utilitas China telah memulai pembangunan beberapa unit AP1000 di dua tapak terpilih untuk rencana operasi pada 2013-2015. AP1000 sebagai desain PWR berdasarkan teknologi teruji dari desain PWR lainnya yang dibuat oleh Westinghouse dengan penguatan pada sistem keselamatan pasif dengan demikian dapat dipertimbangkan untuk dibangun di Indonesia bila mengacu pada persyaratan pada PP 43/2006 mengenai Perijinan Reaktor Nuklir. Namun demikian, desain tersebut perlu diverifikasi oleh Technical Support Organization (TSO) independen sebelum dapat dibangun di Indonesia. Verifikasi dapat dilakukan menggunakan paket program RELAP5 dalam bentuk analisis kecelakaan. Selama ini analisis kecelakaan PLTN dilakukan untuk tipe PWR 1000 MWe dari generasi II atau tipe konvensional. Mengingat saat ini referensi yang menggambarkan teknologi AP1000 yang menyertakan teknologi keselamatan pasif sudah tersedia maka dilakukan kegiatan pemodelan yang nantinya dapat digunakan untuk melakukan analisis kecelakaan. Metode pengembangan model mengacu pada pedoman IAEA yang terdiri dari pengumpulan data instalasi, pengembangan engineering data dan penyusunan input deck, verifikasi dan validasi data input. Model yang berhasil dikembangkan secara umum telah mewakili sistem AP1000 secara keseluruhan dan dianggap sebagai model dasar. Model tersebut telah diverifikasi dan divalidasi dengan data desain yang terdapat pada referensi dimana respon parameter termohidraulika menunjukkan perbedaan hasil ± 3% selain untuk parameter penurunan tekanan teras yang lebih rendah 13%. Sebagai model dasar, input deck yang diperoleh dapat dikembangkan lebih lanjut dengan mengintegrasikan pemodelan sistem keselamatan, sistem proteksi, dan sistem kendali yang spesifik AP1000 untuk keperluan simulasi keselamatan yang lebih rinci.Kata kunci: pemodelan, Generasi III+, RELAP5.   Westinghouse?s AP1000 reactor design is the first Generation III+ nuclear power reactor to receive final design approval from the U.S. Nuclear Regulatory Commission (NRC). Currently, the China?s utilities are starting construction several units of AP1000 on two selected sites for scheduled operation in 2013?2015. The AP1000, based on proven technology of Westinghouse-designed PWR with enhancement on the passive safety system, could be considered to be built in Indonesia referring to the requirements of government regulation No. 43/2006 regarding the Nuclear Reactor Licensing. To be accepted by the regulation agency, the design needs to be verified by independent Technical Support Organization (TSO), which can be done using RELAP5 computer code as accident analyses. Currently, NPP safety accident analysis is performed for PWR 1000 MWe of generation II or conventional type. Considering that nowadays references about the technology of AP1000 that includes passive safety technology has been available and assessed, a modeling activity used for future accident analyzes is introduced. Method for developing the model refers to IAEA guide consisting of plant data collection, engineering data and input deck development, and verification and validation of input data. The model developed should be considered preliminary but has been generally representing the AP1000 systems as the basic model. The model has been verified and validated by comparing thermalhidraulic parameter responses with design data in references with ± 13% deviation except for core pressure drop with 13% lower than design. As a basic model, the input deck is ready for further development by integrating safety system, protection system and control system model specified for AP1000 for purposes of safety simulation in detailed way. Keywords: Modeling, Generation III+ , RELAP5.
IbM KELOMPOK USAHA CAMILAN “SUMBER REJEKI” Suparno, Suparno; Salean, Dantje; Tjahjono, Hendro
JPM17 Vol 1, No 02 (2015)
Publisher : JPM17

Show Abstract | Download Original | Original Source | Check in Google Scholar

Abstract

“Sumber Rejeki” micro owner is Mr.Suwito and Mr. Adi Suprianto which located at Desa Cangkring Kecamatan Ngadirojo Kabupaten Pacitan. This business has been established in 2007 and operated in the field of business Kolong Klitik. The main problems faced through “Sumber Rejeki” micro business are the quantity of production is not optimal yet and both could’nt make financial bookkeeping business and less distribution. By using science and technology from the Ipteks bagi Masyarakat (IbM) can be expected improving the quantity and quality of Kolong Klitik production, as well as to make financial efforts and to expand marketing network.Keywords : Production, Finance and Marketing.
ANALYSIS OF THE EFFECT OF ELEVATION DIFFERENCE BETWEEN HEATER AND COOLER POSITION IN THE FASSIP-01 TEST LOOP USING RELAP5 Ekariansyah, Andi Sofrany; Tjahjono, Hendro; Juarsa, Mulya; Widodo, Surip
SIGMA EPSILON - Buletin Ilmiah Teknologi Keselamatan Reaktor Nuklir Vol 19, No 1 (2015): Februari 2015
Publisher : Badan Tenaga Nuklir Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (1072.747 KB) | DOI: 10.17146/sigma.2015.19.1.2895

Abstract

To understand the natural circulation phenomena on the passive residual heat removal system (PRHRS), development of a test section describing that phenomena in particular in the one phase condition is required. That test facility is named as FASSIP-01 in form of a vertically closed loop consisting of piping compo-nents, one cylinder tank featured with heater elements and one cooler. The heater tank will work as the heat source, and the cooler as the heat sink. This research is intended to support the experimental activity of the FASSIP-01 by conducting a simulation using the RELAP5/SCDAP/Mod3.4. Beside the standard loop configuration, the simulation is also conducted by varying the elevation of heater and cooler position to evaluate the best position resulting in the most optimal natural circulation. The results will be used as the comparison with the later performed experiment. The simulation result shows that for the case where the heater position is at the same level with the cooler position, the temperature distribution of the water after the heater and after the cooler are higher than the other two position. Looking at the natural circulation, that position results in the lowest mass flow. The position with the heater below the cooler will result in the best mass flow. On that position, only an optimiza-tion in the heat transfer surface area is needed to increase the heat transfer coefficient and secondary mass flow to remove the heat are needed to obtain more optimal performance of the water circulation caused by the density difference in the FASSIP-01 test loop.