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Jurnal Teknologi Reaktor Nuklir Tri Dasa Mega

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Published by Badan Tenaga Nuklir Nasional

ISSN : 1411240X EISSN : 25279963 DOI : -

Core Subject : Science,

Energy Materials Science & Nanotechnology

Jurnal Teknologi Reaktor Nuklir "TRI DASA MEGA" adalah forum penulisan ilmiah tentang hasil kajian, penelitian dan pengembangan tentang reaktor nuklir pada umumnya, yang meliputi fisika reaktor, termohidrolika reaktor, teknologi reaktor, instrumentasi reaktor, operasi reaktor dan lain-lain yang menyangkut reaktor nukli. Frekuensi terbit tiga (3) kali setahun setiap bulan Februari, Juni dan Oktober.

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Search results for , issue "Vol 18, No 1 (2016): Februari 2016" : 5 Documents clear

INVESTIGASI KARAKTERISTIK TERMOHIDROLIKA TERAS REAKTOR DAYA KECIL DENGAN PENDINGINAN SIRKULASI ALAM MENGGUNAKAN RELAP5 Susyadi Susyadi; Hendro Tjahjono; Sukmanto Dibyo; Jupiter Sitorus Pane
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 18, No 1 (2016): Februari 2016
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

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 LAJU ALIR MASSA DALAM PURIFIKASI PENDINGIN RGTT200K UNTUK PROSES KONVERSI KARBONMONOKSIDA Sumijanto Sumijanto; Sriyono Sriyono
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 18, No 1 (2016): Februari 2016
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

ABSTRAK OPTIMASI LAJU ALIR MASSA DALAM PURIFIKASI PENDINGIN RGTT200K UNTUK PROSES KONVERSI KARBONMONOKSIDA. Karbonmonoksida adalah spesi yang sulit dipisahkan dari helium pendingin reaktor karena mempunyai ukuran molekul relatif kecil sehingga diperlukan proses konversi menjadi karbondioksida. Laju konversi karbonmonoksida dalam sistem purifikasi dipengaruhi oleh beberapa parameter diantaranya konsentrasi, temperatur dan laju alir massa. Dalam penelitian ini dilakukan optimasi laju alir massa dalam purifikasi pendingin RGTT200K untuk proses konversi karbonmonoksida. Optimasi dilakukan melalui simulasi proses konversi karbonmonoksida menggunakan perangkat lunak Super Pro Designer. Laju pengurangan spesi reaktan, laju pertumbuhan spesi antara dan spesi produk dalam kesetimbangan reaksi konversi dianalisis untuk memperoleh optimasi laju alir massa purifikasi terhadap proses konversi karbonmonoksida. Tujuan penelitian ini adalah menyediakan data laju alir massa purifikasi untuk pembuatan dasar desain sistem purifikasi helium pendingin RGTT200K. Hasil analisis menunjukkan bahwa pada laju alir massa 0,6 kg/detik proses konversi belum optimal, pada laju alir massa 1,2 kg/detik mencapai optimal dan pada laju alir 3,6 kg/detik s/d 12,0 kg/detik tidak efektif. Untuk memdukung dasar desain sistem purifikasi helium pendingin RGTT200K maka laju alir massa purifikasi untuk proses konversi karbonmonoksida digunakan laju alir massa 1,2 kg/detik. Kata kunci: Karbonmonoksida, konversi, purifikasi, laju alir massa, RGTT200K. ABSTRACT OPTIMIZATION OF MASS FLOW RATE IN RGTT200K COOLANT PURIFICATION FOR CARBONMONOXIDE CONVERSION PROCESS. Carbonmonoxide is a species that is difficult to be separated from the reactor coolant helium because it has a relatively small molecular size. So it needs a process of conversion from carbonmonoxide to carbondioxide. The rate of conversion of carbonmonoxide in the purification system is influenced by several parameters including concentration, temperature and mass flow rate. In this research, optimization of the mass flow rate in coolant purification of RGTT200K for carbonmonoxide conversion process was done. Optimization is carried out by using software Super Pro Designer. The rate of reduction of reactant species, the growth rate between the species and the species products in the conversion reactions equilibrium were analyzed to derive the mass flow rate optimization of purification for carbonmonoxide conversion process. The purpose of this study is to find the mass flow rate of purification for the preparation of the basic design of the RGTT200K coolant helium purification system. The analysis showed that the helium mass flow rate of 0.6 kg/second resulted in an unoptimal conversion process. The optimal conversion process was reached at a mass flow rate of 1.2 kg/second. A flow rate of 3.6 kg/second – 12 kg/second resulted in an ineffective process. For supporting the basic design of the RGTT200K helium purification system, the mass flow rate for carbonmonoxide conversion process is suggested to be1.2 kg/second . Keywords: Carbonmonoxide, conversion, purification, mass flow rate, RGTT200K.

NEUTRONICS ANALYSIS ON MINI TEST FUEL IN THE RSG-GAS CORE Tukiran Surbakti; Tagor Malem Sembiring
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 18, No 1 (2016): Februari 2016
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Abstract NEUTRONICS ANALYSIS ON MINI TEST FUEL IN THE RSG-GAS CORE. Research of UMo fuel for research reactor has been developing right now. The fuel of research reactor used is uranium low enrichment with high density. For supporting the development of fuel, an assessment of mini fuel in the RSG-GAS core was performed. The mini fuel are U7Mo-Al and U6Zr-Al with densitis of 7.0gU/cc and 5.2 gU/cc, respectively. The size of both fuel are the same namely 630x70.75x1.30 mm were inserted to the 3 plates of dummy fuel. Before being irradiated in the core, a calculation for safety analysis from neutronics and thermohydrolics aspects were required. However, in this paper will discuss safety analysis of the U7Mo-Al and U6Zr-Al mini fuels from neutronic point of view. The calculation was done using WIMSD-5B and Batan-3DIFF code. The result showed that both of the mini fuels could be irradiated in the RSG-GAS core with burn up less than 70 % within 12 cycles of operation without over limiting the safety margin. Power density of U7Mo-Al mini fuel bigger than U6Zr-Al fuel. Key words: mini fuel, neutronics analysis, reactor core, safety analysis Abstrak ANALISIS NEUTRONIK ELEMEN BAKAR UJI MINI DI TERAS RSG-GAS. Penelitian tentang bahan bakar UMo untuk reaktor riset terus berkembang saat ini. Bahan bakar reaktor riset yang digunakan adalah uranium pengkayaan rendah namun densitas tinggi. Untuk mendukung pengembangan bahan bakar dilakukan uji elemen bakar mini di teras reakror RSG-GAS dengan tujuan menentukan jumlah siklus di dalam teras sehingga tercapai fraksi bakar maksimum. Bahan bakar yang diuji adalah U7Mo-Al dengan densitas 7,0 gU/cc dan U6Zr-Al densitas 5,2 gU/cc. Ukuran kedua bahan bakar uji tersebut adalah sama 630x70,75x1,30 mm dimasukkan masing masing kedalam 3 pelat dummy bahan bakar. Sebelum diiradiasi ke dalam teras reaktor maka perlu dilakukan perhitungan keselamatan baik secara neutronik maupun termohidrolik. Dalam makalah ini akan dibahas analisis keselamatan uji bahan bakar mini U7Mo-Al dan U6Zr-Al ditinjau dari segi neutronik. Perhitungan dilakukan dengan menggunakan program komputer WIMSD-5B dan Batan-3DIFF. Hasil analisis menunjukkan bahwa kedua bahan bakar uji dapat diiradiasi dengan derajat bakar < 70 % selama 12 siklus operasi tanpa melampaui batas keselamatan neutronik. Kerapatan panas bahan bakar uji U7Mo-Al lebih besar dari bahan bakar U6Zr-Al. Kata kunci: Bahan bakar mini, analisis neutronik, teras reaktor, analisis keselamatan

KARAKTERISTIK PERPINDAHAN PANAS KONVEKSI ALAMIAH ALIRAN NANOFLUIDA AL2O3-AIR DI DALAM PIPA ANULUS VERTIKAL Reinaldy Nazar
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 18, No 1 (2016): Februari 2016
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

ABSTRAK KARAKTERISTIK PERPINDAHAN PANAS KONVEKSI ALAMIAH ALIRAN NANOFLUIDA AL2O3-AIR DI DALAM PIPA ANULUS VERTIKAL. Hasil beberapa penelitian menunjukan bahwa nanofluida memiliki karakteristik termal yang lebih baik dibandingkan dengan fluida konvensional (air). Berkaitan dengan hal tersebut, saat ini sedang berkembang pemikiran untuk menggunakan nanofluida sebagai fluida perpindahan panas alternatif pada sistem pedingin reaktor. Sementara itu, konveksi alamiah di dalam pipa anulus vertikal merupakan salah satu mekanisme perpindahan panas yang penting dan banyak ditemukan pada reaktor riset TRIGA, reaktor daya generasi baru dan alat konversi energi lainnya. Namun disisi lain karakteristik perpindahan panas nanofluida di dalam pipa anulus vertikal belum banyak diketahui. Oleh karena itu penting dilakukan secara berkesinambungan penelitian-penelitian untuk menganalisis perpindahan panas nanofluida di dalam pipa anulus vertikal. Pada penelitian telah dilakukan analisis numerik menggunakan program computer CFD (computational of fluids dynamic) terhadap karakteristik perpindahan panas konveksi alamiah aliran nanofluida Al2O3-air konsentrasi 2% volume di dalam pipa anulus vertikal. Hasil kajian ini menunjukkan terjadi peningkatan kinerja perpindahan panas (bilangan Nuselt- NU) sebesar 20,5% - 35%. Pada moda konveksi alamiah dengan bilangan 2,4708e+09 £ Ra £ 1,9554e+13 diperoleh korelasi empirik untuk air adalah dan korelasi empirik untuk nanofluida Al2O3-air adalah Kata kunci: Nanofluida Al2O3-air, konveksi alamiah, pipa anulus vertikal ABSTRACT THE CHARACTERISTICS OF NATURAL CONVECTIVE HEAT TRANSFER OF AL2O3–WATER NANOFLUIDS FLOW IN A VERTICAL ANNULUS PIPE. Results of several research have shown that nanofluids have better thermal characteristics compared to conventional fluid (water). In this regard, currently developing ideas for using nanofluids as an alternative heat transfer fluid in the reactor coolant system. Meanwhile the natural convection in a vertical annulus pipe is one of the important mechanisms of heat transfer and is found at the TRIGA research reactor, the new generation nuclear power plants and other energy conversion devices. On the other hand the heat transfer characteristics of nanofluids in a vertical annulus pipe has not been known. Therefore, it is important to do research continuously to analyze the heat transfer nanofluids in a vertical annulus pipe. In the research has been carried out numerical analysis by using computer code of CFD (computational of fluids dynamic) on natural convection heat transfer characteristics of nanofluids flow of Al2O3-water 2% volume in the vertical annulus pipe. The results showed an increase in heat transfer performance (Nusselt numbers - NU) by 20.5% - 35%. In natural convection mode with Rayleigh numbers 2.4708e+09 £ Ra £ 1.9554e+13 obtained empirical correlations for water is and empirical correlations for Al2O3-water nanofluids is . Keywords: Al2O3-water nanofluids, the natural convection, the vertical annulus pipe

VALIDATION OF SIMBAT-PWR USING STANDARD CODE OF COBRA-EN ON REACTOR TRANSIENT CONDITION Muhammad Darwis Isnaini; Muhammad Subekti
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 18, No 1 (2016): Februari 2016
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

The validation of Pressurized Water Reactor typed Nuclear Power Plant simulator developed by BATAN (SIMBAT-PWR) using standard code of COBRA-EN on reactor transient condition has been done. The development of SIMBAT-PWR has accomplished several neutronics and thermal-hydraulic calculation modules. Therefore, the validation of the simulator is needed, especially in transient reactor operation condition. The research purpose is for characterizing the thermal-hydraulic parameters of PWR1000 core, which be able to be applied or as a comparison in developing the SIMBAT-PWR. The validation involves the calculation of the thermal-hydraulic parameters using COBRA-EN code. Furthermore, the calculation schemes are based on COBRA-EN with fixed material properties and dynamic properties that calculated by MATPRO subroutine (COBRA-EN+MATPRO) for reactor condition of startup, power rise and power fluctuation from nominal to over power. The comparison of the temperature distribution at nominal 100% power shows that the fuel centerline temperature calculated by SIMBAT-PWR has 8.76% higher result than COBRA-EN result and 7.70% lower result than COBRA-EN+MATPRO. In general, SIMBAT-PWR calculation results on fuel temperature distribution are mostly between COBRA-EN and COBRA-EN+MATPRO results. The deviations of the fuel centerline, fuel surface, inner and outer cladding as well as coolant bulk temperature in the SIMBAT-PWR and the COBRA-EN calculation, are due to the value difference of the gap heat transfer coefficient and the cladding thermal conductivity.

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Pusat Teknologi dan Keselamatan Reaktor Nukir (PTKRN) Badan Tenaga Nuklir Nasional (BATAN) Gedung 80 Kawasan Puspiptek Setu - Tangerang Selatan Banten - Indonesia (15310)

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Kota adm. jakarta selatan,

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