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Ganendra: Majalah IPTEK Nuklir
Published by Badan Tenaga Nuklir Nasional
ISSN : 14106957     EISSN : 25035029     DOI : https://doi.org/10.17146/gnd
Core Subject : Science, Education,
Computer Science & IT Other
Jurnal Iptek Nuklir Ganendra merupakan jurnal ilmiah hasil litbang dalam bidang iptek nuklir, diterbitkan oleh Pusat Teknologi Akselerator dan Proses Bahan (PTAPB) - BATAN Yogyakarta. Frekuensi terbit dua kali setahun setiap bulan Januari dan Juli.
Arjuna Subject : -
Articles 5 Documents
 1 
Search results for , issue "Volume 24 Nomor 1 Januari 2021" : 5 Documents clear
ANALISIS AKURASI PAKET PROGRAM WIMSD-5B/CITATION DALAM PERHITUNGAN KRITIKALITAS REAKTOR MSRE ACCURATION ANALYSIS OF THE WIMSD-5B/CITATION CODES ON THE CRITICALITY CALCULATION OF THE MSRE REACTOR S. Permana; D. Tamaza; H. Sid'qon
GANENDRA Majalah IPTEK Nuklir Volume 24 Nomor 1 Januari 2021
Publisher : Website

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/gnd.2021.24.1.6146

Abstract

ANALISIS AKURASI PAKET PROGRAM WIMSD-5B/CITATION DALAM PERHITUNGAN KRITIKALITAS REAKTOR MSRE. Paket program WIMSD-5B/CITATION telah digunakan sebagai perangkat analitik kekritisan berbagai jenis reaktor dan memberikan hasil yang memuaskan.  Meskipun demikian, paket program ini belum pernah dilakukan untuk menganalisis reaktor MSR (Molten Salt Reactor). Penelitian ini menyajikan analisis akurasi WIMSD-5B/CITATION untuk kritikalitas reaktor MSRE (Molten Salt Reactor Experiment), yaitu reaktor MSR yang pernah dioperasikan sebagai fasilitas eksperimen.  Tujuan penelitian mengetahui akurasi WIMSD-5B/CITATION untuk kritikalitas bahan bakar Tipe C di reaktor MSRE.  Kritikalitas teras reaktor MSRE dihitung dengan 6 kelompok energi neutron dengan model geometri R-Z.  Hasil perhitungan menunjukkan model sel yang menempatkan celah di tengah grafit (Model 1) lebih baik dibanding dengan sel yang menempatkan celah bahan bakar diluar (Model 2). Namun demikian perbedaan relatif dengan eksperimen masih tinggi karena ada perbedaan relatif 7,23%. Akurasi perhitungan kritikalitas didominasi oleh faktor model geometri sel dan teras.  Kemudian data jumlah void dan komposisi pengotor Li-6 juga memiliki pengaruh yang signifikan. Hasil penelitian juga menunjukkan bahwa fluks neutron dan faktor puncak daya radial di teras MSRE sangat sensitif dengan model sel bahan bakar.
A NOVEL DESIGN OF 17.5 KV HV FEEDTHROUGH FOR ARJUNA 2.0 Saefurrochman Saefurrochman; Darsono Darsono; Suhadah Rabi’atul Adabiah; Elin Nuraini; Sutadi Sutadi; Tanti Ardiyati
GANENDRA Majalah IPTEK Nuklir Volume 24 Nomor 1 Januari 2021
Publisher : Website

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/gnd.2021.24.1.6223

Abstract

A NOVEL DESIGN OF 17.5 KV HV FEEDTHROUGH FOR ARJUNA 2.0. A novel design of the 17.5 kV feedthrough for Arjuna 2.0 Cockcroft Walton generator has been proposed. It is used for connecting the output of RF transformer oscillator (in the outside of horizontal vessel) with the input of voltage multiplier (inside of horizontal vessel) of the Cockcroft Walton generator. It was equipped by covers on left and right side. The designed feedthrough was simple, compact, easy to manufacture, high performance to prevent flashover and also it was applied to Arjuna 2.0 Cockcroft Walton. It was made from teflon (PTFE) and solid copper, which have high dielectric strength, capable of withstanding press loads, and easy to manufacture. The shortest distance between grounding with conductor radially was 43.25 mm, and 253.5 mm for feedthrough surface. The design was verified by Finite Element Method software and continued with performance testing. According to simulation, the stress of voltage is high about 16 kV to 17.5 kV on feedthrough conductor and 0 to 3 kV on feedthrough flange. The electric field of the covered feedthrough is lower than the coverless feedthrough. The highest and lowest electric fields are 1.26 x 106 V/m and 1 x 105 to 2 x 105 V/m respectively. Furthermore, feedthrough has been tested up to 120 kV and no discharge occurred. It means this design can be operated for 17.5 kV and it was successful installed on Arjuna 2.0 Cockcroft Walton generator.
CYLINDRICAL SHELL ANALYSIS OF REACTOR PRESSURE VESSEL FOR RDE Sri Sudadiyo
GANENDRA Majalah IPTEK Nuklir Volume 24 Nomor 1 Januari 2021
Publisher : Website

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/gnd.2021.24.1.5191

Abstract

CYLINDRICAL SHELL ANALYSIS OF REACTOR PRESSURE VESSEL FOR RDE. The present study deals with the design process analysis of cylindrical shell for Reactor Pressure Vessel (RPV) of Reaktor Daya Eksperimental (RDE). The RDE is prepared by BATAN for nuclear technology provider in Indonesia. RPV is a container for confining helium gas at elevated pressure and temperature (circa 700 °C). In RPV operation, mechanical stresses act as in consequence of internal pressure (3 MPa), external pressure, and different loads due to dead weight and helium content load. Therefore, if the RPV could not retain its material strength or exceed the maximum allowable shear stress it will cause failure. The applications and validity of Fortran code (RPV_RDE.exe) for the design analysis are represented by two simulation cases, which indicate good calculation results of design outputs compared to analytic solutions. Design outputs have met the safe requirements for the minimum wall thickness of cylindrical shell in upper portion of 60 mm and in lower portion of 100 mm, respectively.
COMPARISON OF MEASURED AND CALCULATED CONTROL ROD REACTIVITY OF THE RSG-GAS CORE T. Surbakti; W. Luthfi; Purwadi Purwadi; D. Hartanto
GANENDRA Majalah IPTEK Nuklir Volume 24 Nomor 1 Januari 2021
Publisher : Website

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/gnd.2021.24.1.5948

Abstract

COMPARISON OF MEASURED AND CALCULATED CONTROL ROD REACTIVITY OF THE RSG-GAS CORE. The reactivity worth and a calibration curve of control rods are important physical parameters on account of the nuclear safety of the RSG-GAS reactor core. Reactivity characteristics of control rods must be determined not only, but also in the first criticality and also after any substantial changes in the reactor core. The required time to measure the reactivity worth by various methods depends mostly on the quality of the reactivity measurement devices. The rod-insertion method has some specific advantages which make such compensation unnecessary while yielding both integral and differential reactivity worth curves, as well as total reactivity worth. In this study, the total control rod worth of RSG-GAS is calibrated experimentally using a rod-insertion method to verify calculation results for the new core configuration. Calculations were done by diffusion and Monte Carlo methods using Batan-3DIFF and MCNP5 codes. Total control rod reactivity worth of the control rod is obtained 17.54 $,  17.03 $, and 17.87 $ by experiment,  Batan-3DIFF, and MCNP5 calculations, respectively. The relative difference between the experimental and calculated values of control rod reactivity worth is about 3.0 %, which indicates a good agreement between the applied experimental method and calculation.
STUDI PENGARUH PERBEDAAN KETEBALAN FREEZE-VALVE DI MSR (MOLTEN SALT REACTOR) DALAM PENGENDALIAN KECELAKAAN Virgo Eben E. M; Mustari A. P. A.; Irwanto D.; Permana S.; Pramuditya S.
GANENDRA Majalah IPTEK Nuklir Volume 24 Nomor 1 Januari 2021
Publisher : Website

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/gnd.2021.24.1.5957

Abstract

STUDI PENGARUH PERBEDAAN KETEBALAN FREEZE-VALVE DI MSR (MOLTEN SALT REACTOR) DALAM PENGENDALIAN KECELAKAAN. MSR merupakan sebuah tipe reaktor yang menggunakan bahan bakar cair, yakni garam cair yang sekaligus berfungsi sebagai pendinginnya. Meskipun MSR diakui unggul dari segi keamanan dan menjadi salah satu kandidat reaktor Generasi IV, namun dibutuhkan penelitian lebih lanjut untuk mencegah kecelakaan reaktor akibat peningkatan suhu bahan bakar cair. Sebuah freeze-valve merupakan salah satu sistem keamanan yang sangat berfungsi di MSR. Sebuah freeze-valve didesain untuk meleleh saat suhu bahan bakar mendekati titik leleh dinding reaktor dan membuka jalan bahan bakar menuju subcritical-tank. Sebuah eksperimen sederhana telah berhasil dilakukan pada penelitian ini untuk mempelajari mekanisme kerja freeze-valve tersebut. Penelitian ini dilakukan untuk menganalisa pengaruh perbedaan ketebalan pada freeze-valve (parafin) yang dialiri fluida panas bersuhu 800C. Variasi ketebalan yang digunakan adalah 10, 13, dan 20 mm dengan diameter yang sama, yaitu 23 mm. Berdasarkan variasi ketebalan tersebut ditemukan bahwa jika keadaan kecelakaan didesain pada saat suhu fluida di dalam tabung mencapai 800C, maka dibutuhkan sebuah freeze-valve dengan tebal 7 mm agar dapat terbuka dalam waktu 10 menit dan fluida dapat mengalir menuju subcritical tank.

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