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BORIC ACID RADIOLYSIS IN PRIMARY COOLANT WATER OF PWR AT TEMPERATURE OF 250OC Sunaryo, Geni Rina
Jurnal Pengembangan Energi Nuklir Vol 19, No 1 (2017): Juni 2017
Publisher : Pusat Kajian Sistem Energi Nuklir, Badan Tenaga Nuklir Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/jpen.2017.19.1.3192

Abstract

BORIC ACID RADIOLYSIS IN PRIMARY COOLANT WATER OF PWR AT TEMPERATURE OF 250oC. The existence of oxygen in the primary coolant system of PWR could lead to corrosion, hence it is very important to suppress the oxygen concentration in the system. Therefore, study of the effect of boric acid addition into the primary coolant water system of PWR to suppress oxygen concentration resulted from gamma-ray radiation is essential to be performed. The aim of this research is to understand reaction mechanism at temperature of 2500C and the effect of boric acid adding toward oxygen concentration in the PWR primary coolant water. Methodology used is simulation using Facsimile software. Input for the software namely radiolysis reaction mechanism for pure water, G value from radiolysis product, dose rate of 1 and 104 Gy/s, aeration and deaeration system, and specific reaction of boric acid with hydroxyl radical and hydrated electron at temperature 250C and 3000C. The output are in the form of irradiation time vs oxygen concentration time series. The results show that the oxygen production increase significantly with the irradiation time and reach the saturated concentration at 107s. Based on the plot of oxygen?s concentration at 107s vs boric acid, several results are as following: oxygen concentration significantly suppressed by boric acid addition and gives the exponential decreasement, the higher dose rate gives the higher concentration of oxygen, the aeration system gives no effect on suppressing oxygen concentration at boric acid addition up to 0.1M.
APLIKASI MSC PATRAN UNTUK PENENTUAN RENTANG MAKSIMUM PENYANGGA PIPA PRIMER REAKTOR AP1000 Saragi, Elfrida; Hafid, Abdul; Sunaryo, Geni Rina
Jurnal Pengembangan Energi Nuklir Vol 17, No 1 (2015): Juni 2015
Publisher : Pusat Kajian Sistem Energi Nuklir, Badan Tenaga Nuklir Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (1276.142 KB) | DOI: 10.17146/jpen.2015.17.1.2613

Abstract

ABSTRAK APLIKASI MSC PATRAN UNTUK PENENTUAN RENTANG MAKSIMUM PENYANGGA PIPA PRIMER REAKTOR AP1000. Penyangga pipa digunakan antara lain untuk menjaga agar pipa tidak membebani komponen dan mencegah terjadinya lendutan yang berlebihan. Penentuan posisi penyangga pipa ditetapkan oleh beberapa faktor, seperti adanya katup, adanya belokan pipa dan jarak antara dua komponen utama reaktor yaitu tangki reaktor dan pembangkit uap.Untuk transpor panas dari tangki reaktor ke pembangkit uap digunakan pipa hotleg. Tujuan penelitian ini adalah untuk dapat menentukan batas jarak penyangga yang baik dan sudut belok pipa pada pipa primer reaktor daya AP1000 berdiameter 37,5inchi diameter luar dan 31 inchi diameter dalam. Metode analisis yang digunakan adalah metode komputasi dengan pemodelan menggunakan software MSC Patran.Hasil perhitungan menunjukkan bahwa semakin jauh jarak penyangga pipa maka besar lendutan yang terjadi makin besar.Nilai maksimal yang cukup baik dan sesuai standar ASME adalah pada jarak 5 m dengan sudut belok pipa 45 derajat.Pada jarak tersebut defleksi maksimumyang terjadi sebesar 1.76 cm dan tegangan tekuk sebesar 2.06 MPa. Kata kunci: Tegangan tekuk, Defleksi, Penyangga pipa, Hotlegreaktor AP1000. ABSTRACT APPLICATION OF MSC-PATRANTO DETERMINE THE MAXIMUM RANGE SUPPORT OF PRIMARY PIPES NUCLEAR REACTOR AP1000. Pipe supports used among others, to keep the pipes from overloding the components and prevent excessive deflection. The position of the pipe support is determined by several factors, such as the presence of valves, pipe bends and the distance between the two main components of reactor. Heat transport from reactor tank to the steam generatorare performed using hotleg pipe. The purpose of this study was to determine a safe support distance limit and the angle of the pipe turn and bendingon the primary pipe of AP1000 power reactor with the outer pipe diameter of 37.5 inches, and the inner diameter of the pipe is 31 inches.The analytical method used is the computational modeling methodsusing the MSC Patran software. The calculation resultsshow that the greater the distance of the pipe support, then deflection occurs is greater. The maximum value that is quite good, andin accord to ASME standards is at a distance of 5 meter and the angle of pipe turn is 45 degree. At that distance, the maximum deflection occurs is 1.76 cm and bending stress is 2.06 MPa. Keywords: Deflection, Bending stress, Support pipes, Hotleg reactor AP1000.
MEKANISME REAKSI ASAM BORAT DENGAN PRODUK RADIOLISIS AKIBAT RADIASI SINAR- PADA TEMPERATUR 25OC Sunaryo, Geni Rina
Jurnal Pengembangan Energi Nuklir Vol 14, No 2 (2012): Desember 2012
Publisher : Pusat Kajian Sistem Energi Nuklir, Badan Tenaga Nuklir Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (818.301 KB) | DOI: 10.17146/jpen.2012.14.2.1479

Abstract

ABSTRAKMEKANISME REAKSI ASAM BORAT DENGAN PRODUK RADIOLISIS AKIBAT RADIASI SINAR-? PADA TEMPERATUR 25OC. Telah dilakukan simulasi yang bertujuan untuk memahami mekanisme reaksi antara asam borat (H3BO3) yang ditambahkan kedalam air pendingin primer PWR dengan produk radiolisis akibat radiasi dengan sinar-? pada temperatur 25oC. Simulasi dilakukan dengan menggunakan perangkat lunak ?Facsimile? yang berbasis kinetika reaksi yang berkelanjutan. Sebagai masukan adalah set reaksi kimia yang terdiri dari 61 jenis reaksi dengan konstanta kecepatan reaksinya, nilai-G spesi radiolisis akibat radiasi sinar-?, laju dosis 10 dan 104 Gy/s, konsentrasi awal oksigen yang berhubungan dengan sistem aerasi (0,25M), deaerasi dan konsentrasi asam borat hingga konsentrasi 1M. Luaran di program berupa seri perubahan konsentrasi vs waktu iradiasi. Data luaran kemudian diolah menggunakan perangkat pembuat grafik ?Origin?. Validasi dilakukan dengan membandingkannya dengan hasil simulasi sebelumnya. Hasil validasi menunjukkan perbedaan yang tidak signifikan, sehingga diputuskan bahwa set reaksi sekarang adalah valid. Penambahan asam borat menekan konsentrasi oksigen secara signifikan. Hubungan kenaikan logaritmik penambahan konsentrasi H3BO3 vs produk oksigen menunjukkan hubungan linear yang menurun. Dari hasil simulasi dapat dipahami bahwa penambahan H3BO3 tidak hanya mengatur reaktivitas neutron pada temperatur 25oC tetapi juga memberikan imbas positif didalam menekan konsentrasi produk oksigen yang memegang peran penting di dalam proses korosi.Kata kunci: radiolisis, sinar-?, larutan H3BO3, facsimile ABSTRACTTHE EFFECT OF BORIC ACID ON OXYGEN SUPPRESSING UNDER ??RAY IRRADIATION AT 25OC. Simulation to understanding the reaction mechanism between boric acid that is being added into the PWR primary water and radiolysis products under ?-rays irradiation at 25oC was done. Simulation has been done by using ?Facsimile? software based on continuing kinetic reaction. As inputs are set reactions that consist of 61 reactions, G-values under ??rays irradiation, doserate of 10 and 104 Gy/s, initial concentration of oxygen for aeration (0.25M) and deaeration, and boric acid up to 1M. Outputs are series of concentration vs irradiation time. The putput data is being analysed by plotting them into graph by using ?Origin?. Validation was done by comparing the results with the previous work. From validation, it is know that the set reaction that is being used does not give any significant difference, then dicided that the set reactions used is valid. The relation between concentration of boric acid and oxygen concentration logarithmically is linearly decrease. From the simulation, it can be understood either that the addition of H3BO3 is not only for controlling the neutron reactivity but also give positive effect on suppressing the oxygen concentration that play role on corrosion process.Keywords: radiolyses, ?-ray, H3BO3, facsimile
VERIFICATION TO THE RSG-GAS FUEL DISCHARGE BURN-UP USING SRAC2006 MODULE OF COREBN/HIST Susilo, Jati; Sembiring, Tagor Malem; Imron, M; Sunaryo, Geni Rina
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 20, No 1 (2018): Februari 2018
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (4760.282 KB) | DOI: 10.17146/tdm.2018.20.1.4041

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For 30 years operation, some of the modifications to the RSG GAS core has been done, that are changes included the type of fuel from U3O8-Al to U3Si2-Al with the same density 2.96 gU/cc, the loading pattern of standard fuel elements/fuel control elements from 6/1 & 6/2 to 5/1 pattern, and in core fuel management calculation tool has been change from IAFUEL to BATAN-FUEL. To obtain an extension of the operating license for the next 10 years, the RSG-GAS Periodic Safety Assessment Document is need to prepared. According to the Regulatory Body Chairman Regulation No. 2 2015, RSG-GAS safety assessment should be done independently. As part of this assessment the fuel discharge burn-up must be estimated. In this research, to ensure that the misposition of fuel element in the core has not occurred, the investigation to the document operating report related the fuel placement has been done. Therefore, by using 78th to 93rd operation data, verify of the fuel discharge burn-up of the RSG-GAS has been performed by using SRAC2006 module of COREBN/HIST. In addition, the results of these calculations are also made comparative with the operating report data that is calculated by using BATAN-FUEL. Maximum fuel discharge burn-up (57.73% of U-235) was verified still under permissible value determined by the regulatory body (<60% of U-235). Maximum differences value between two computer codes was about 2.12 % of U-235 (3.80%) that is fuel at the B-7 position. Fuel discharge burn-up of RSG-GAS showed almost the same value for each the operation cycle, range of 1.52% of U-235. So it can be concluded that the RSG-GAS core operation over the last ten years was in good fuel management performance, in accordance with the design. BATAN-FUEL has been comformed well enough with COREBN/HIST. Keywords: Discharge Burn-Up, RSG-GAS, COREBN/HIST, BATAN-FUEL Verifikasi Terhadap Burn-Up Buang Bahan Bakar Teras RSG-GAS Menggunakan SRAC2006 Modul COREBN/HIST. Selama 30 tahun beroperasi, RSG-GAS telah mengalami perubahan modifikasi antara lain jenis bahan bakar dari U3O8-Al menjadi U3Si2-Al dengan kerapatan sama 2,96 gU/cc, pola pemuatan bahan bakar standar/elemen kendali dari pola 6/1 & 6/2 menjadi pola 5/1, dan alat perhitungan manajemen bahan bakar IAFUEL dengan BATAN-FUEL. Untuk memperoleh perpanjangan ijin operasi selama 10 tahun ke depan, maka perlu disiapkan dokumen Penilaian Keselamatan Berkala RSG-GAS. Berdasarkan PerKa BAPETEN No. 2 Tahun 2015, maka penilaian keselamatan RSG-GAS harus dilakukan secara independen. Salah satu parameter yang perlu diverifikasi adalah nilai bahan bakar buang. Dalam penelitian ini, dilakukan investigasi terhadap dokumen Laporan Operasi untuk memastikan bahwa tidak terjadi kesalahan penempatan bahan bakar. Selanjutnya, berdasarkan data siklus operasi teras ke 78 sampai dengan 93, dilakukan verifikasi nilai burn-up buang bahan bakar RSG-GAS dengan menggunakan SRAC2006 modul COREBN/HIST. Selain itu, hasil perhitungan tersebut juga dilakukan komparasi dengan data laporan operasi yaitu data hasil perhitungan menggunakan BATAN?FUEL. Fraksi bakar buang bahan bakar terbesar (57,73% U-235) terverivikasi masih di bawah nilai limit yang ditetapkan oleh badan pengawas (<60% U-235). Perbedaan hasil perhitungan terbesar kedua program computer sebesar 2,12% U-235 (3,80%) yaitu pada posisi B-7. Fluktuasi burn-up buang bahan bakar menunjukkan nilai yang hampir sama untuk tiap-tiap siklus operasi, jarak (range) sebesar 1,52% U-235. Sehingga dapat disimpulkan bahwa operasi teras RSG-GAS selama sepuluh tahun terakhir menunjukkan performa manajemen bahan bakar yang baik, sesuai desain. BATAN-FUEL telah terkonfirmasi cukup baik dengan COREBN/HIST. Kata kunci: Burn-up buang, RSG-GAS, COREBN/HIST, BATAN-FUEL
PEMILIHAN TEKNOLOGI DESALINASI NUKLIR DI PROVINSI KALIMANTAN TIMUR Alimah, Siti; Ariyanto, Sudi; Dewita, Erlan; Budiarto, Budiarto; Sunaryo, Geni Rina
Jurnal Pengembangan Energi Nuklir Vol 11, No 1 (2009): Juni 2009
Publisher : Pusat Kajian Sistem Energi Nuklir, Badan Tenaga Nuklir Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (727.787 KB) | DOI: 10.17146/jpen.2009.11.1.1427

Abstract

ABSTRAKPEMILIHAN TEKNOLOGI DESALINASI NUKLIR DI PROVINSI KALIMANTAN TIMUR. Saat ini, kebutuhan listrik di Kalimantan Timur (Kaltim) meningkat dengan kecepatan 12% per tahun. Karena pasokan listrik yang dihasilkan PT. PLN meningkat 8,5% per tahun, maka mengakibatkan terjadinya krisis listrik di daerah tersebut. PLTN (Pembangkit Listrik Tenaga Nuklir) dapat merupakan salah satu alternatif untuk mengatasi masalah tersebut. Demikian pula halnya dengan ketersediaan air bersih, pasokannya juga lebih sedikit dari pada kebutuhan. Sehingga memerlukan suatu upaya yang serius. Desalinasi nuklir adalah suatu proses memisahkan garam terlarut dari air laut atau air payau, yang dikopel dengan PLTN untuk memproduksi air bersih. Ada beberapa jenis teknologi desalinasi yang umum digunakan di dunia, diantaranya MSF (Multi-Stage Flash Distillation), MED (Multi-Effect Distillation) and RO (Reverse Osmosis). Makalah ini menyajikan hasil studi suatu pemilihan teknologi desalinasi untuk memperoleh solusi yang optimal. Pemilihan dilakukan berdasarkan pada pertimbangan 13 parameter penting yang diperkirakan sangat berpengaruh dalam menentukan pemilihan teknologi desalinasi nuklir dengan faktor pembobotan yang mempunyai kisaran 1 sampai 4. Teknologi yang dipilih adalah yang mempunyai nilai bobot paling tinggi. Hasil studi memperlihatkan bahwa MED mempunyai nilai bobot yang paling tinggi yaitu 39, diikuti nilai bobot 36 untuk RO dan 33 untuk MSF. Karena diperlukan air dengan kualitas sekitar 1 ppm untuk memasok PLTN dan untuk memasok kebutuhan masyarakat diperlukan air dengan kualitas dibawah 1000 ppm maka sistem hibrid MED-RO merupakan pilihan optimum untuk memproduksi air bersih.Kata kunci : teknologi desalinasi, MSF, MED, RO, air bersih, kopel, PLTN ABSTRACTSELECTION OF NUCLEAR DESALINATION TECHNOLOGY IN EAST KALIMANTAN PROVINCE. Nowdays, electricity demand in East Kalimantan increases with a rate of 12% per annum. Since the electricity supply produced by PT.PLN increases 8,5% per annum, then it can consequently an occurrence of electricity shortage in the region. NPP may be regarded as one viable option to overcome the problem. In case of fresh water availability, the supply is also less than the demand. Therefore, a serious effort is necessary. Nuclear desalination, which is a process of separating dissolved salts of seawater or brackish water, can be coupled to the NPP to produce fresh water. There are some desalination technology commonly used in the world i.e.MSF (Multi-Stage Flash Distillation), MED (Multi-Effect Distillation) and RO (Reverse Osmosis). This paper shows the study result of selection for desalination technology to obtain the optimum solution. The selection is done based on the thirteen important parameters, which are estimated to affect on determine technology option on the nuclear desalination with a weighing factor with ranges from 1 to 4. The most favourable technology is that with the highest point. The result show that MED has highest weighing factor that is 39, followed 36 for RO and 33 for MSF. Since the water quality requirement to supply NPP is about 1 ppm and to supply public demand is below 1000 ppm, so a hybrid system of MED-RO is optimum option to produce fresh water.Keywords: desalination technology, MSF, MED, RO, fresh water, coupling, NPP
TEMPERATURE DEPENDENCE OF PRIMARY SPECIES G(VALUES) FORMED FROM RADIOLYSIS OF WATER BY INTERACTION OF TRITIUM β-PARTICLES Butarbutar, Sofia Loren; Sriyono, Sriyono; Sunaryo, Geni Rina
Jurnal Pengembangan Energi Nuklir Vol 19, No 1 (2017): Juni 2017
Publisher : Pusat Kajian Sistem Energi Nuklir, Badan Tenaga Nuklir Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (958.406 KB) | DOI: 10.17146/jpen.2017.19.1.3134

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TEMPERATURE DEPENDENCE OF PRIMARY SPECIES G(VALUES) FORMED FROM RADIOLYSIS OF WATER BY INTERACTION OF TRITIUM ?-PARTICLES. G(values) are important to understand the effect of radiolysis of Nuclear Power Plant (NPP) cooling water. Since direct measurements are difficult, hence modeling and computer simulation were carried out to predict radiation chemistry in and around reactor core. G(values) are required to calculate the radiation chemistry. Monte Carlo simulations were used to calculate the G(values) of primary species , H?, H2, ?OH dan H2O2 formed from the radiolysis of tritium ? low energy electron. These radiolytic products can degrade the reactor components and cause corrosion under the reactor operating conditions. G(values) prediction can indirectly contribute to maintain the material reliability. G(values) were calculated at 10-8, 10-7, 10-6 and 10-5 s after ionization at temperature ranges. The calculation were compared with the G(values) of g-ray 60Co. The work aimed to understand temperature effect on the water radiolysis mechanism by the tritium ? electron. The results show that the trend similarity was found on the temperature dependence of G(values) of tritium ? electron and g-ray 60Co. For tritium ? electron, G(values) for free radical were lower than g-ray 60Co, but higher for molecular products as temperature raise at 10-8 and 10-7. The significant differences for these two type of radiations were on G(H2), G(?OH) and G(H?) at 10-6and 10-5 s above 200 oC.
ANALYSIS ON ALMG2 AS RSG-GAS CLADDING MATERIAL CORROSION IN CHLORIDE CONTAINING WATER Bahar, Febrianto; Sriyono, Sriyono; Sunaryo, Geni Rina
GANENDRA Majalah IPTEK Nuklir Volume 21 Nomor 2 Juli 2018
Publisher : PSTA BATAN

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (1382.298 KB) | DOI: 10.17146/gnd.2018.21.2.4271

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AnalYsis On AlMg2 AS RSG-GAS CLADDING material corrosion IN CHLORIDE CONTAINING WATER. The AlMg2is one of an alluminium alloy that used as cladding material for the RSG GA. Siwabessy (RSG-GAS) research reactor in Serpong, Indonesia. The reactor uses demineralized water as primary coolant with 6.5 to 7.5 of pH. A poor treatment of water in primary coolant can lead to the problem of AlMg2 integrity. The primary coolant concentration of chloride must lower than 0.0094 ppm to protect cladding corrosion. The purpose of this study is to determine the effect of temperature and chloride ion concentration to AlMg2. The method in this research is to observe the corrosion rate for AlMg2 material by using Potentiostat. The laboratory experiments were conducted in various temperatures (28, 35, 40 and 45°C) and concentration of sodium chloride of 0.005, 0.010, 0.015, 0.020, 0.025, 0.030 and 0.035 ppm. The results show the corrosion rates were very small, and the highest corrosion rate occurred is 1.23 x 10-3mpy in 0.035 ppm of NaCl at 45°C .
STUDI RADIOLISIS AIR RINGAN DAN PENGUKURAN LAJU DOSIS BAHAN BAKAR TERHADAP JARAK SUMBER RADIASI PADA KOLAM PENYIMPANAN BAHAN BAKAR BEKAS (ISSF) Agustin, Cyntia; Romli, M; Butar-butar, Sofia Loren; Kusumastuti, Rahayu; Sriyono, Sriyono; Sunaryo, Geni Rina
SIGMA EPSILON - Buletin Ilmiah Teknologi Keselamatan Reaktor Nuklir Vol 22, No 2 (2018): November 2018
Publisher : Badan Tenaga Nuklir Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (541.512 KB) | DOI: 10.17146/sigma.2018.22.2.4488

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Studi Literatur Radiolisis Air Ringan dan Pengukuran Laju Dosis Bahan Bakar Terhadap Jarak Sumber Radiasi  Pada Kolam Penyimpanan Bahan Bakar Bekas (ISSF). . Elemen bahan bakar bekas masih mengandung sejumlah uranium diperkaya dengan paparan radiasi yang sangat tinggi, sehingga digunakan air sebagai media penyimpanan bahan bakar bekas pada kolam ISSF agar paparan radiasi bahan bakar tidak keluar ke lingkungan.Paparan radiasi dalam air dapat menyebabkan adanya pembentukan oksidator yang dapat menyebabkan korosi pada material bahan ISSF. Laju dosis dapat terukur dalam suatu sumber radiasi terhadap besarnya penahan radiasi. Laju dosis ini digunakkan sebagai input parameter untuk reaksi radiolysis sehingga konsentrasi pembentukan oksidator dalam air dapat diprediksi. Hubungan antara laju dosis teradap jarak sumber radiasi (tebal penahan) menjadi penting untuk penerapan proteksi radiasi. Metode untuk mengukur laju dosis pada kolam ISSF dilakukan pada rak bahan bakar bekas serta uji cicip pada sebuah kelongsong bahan bakar bekas. Laju dosis diukur dengan detector radiagem dengan kabel yang terbungkus plastik. Data hasil percobaan didapatkan bahwa hubungan antara laju dosis radiasi terhadap sumber radiasi yaitu semakin besar jarak detektor terhadap sumber radiasi semakin kecil laju dosis yang terukur dan bersifat eksponensial.Kata Kunci : Kolam ISSF, radiasi, radiolysis air, laju dosis, detector
WATER CHEMISTRY ANALYSIS IN RSG-GAS SECONDARY COOLING SYSTEM Kusumastuti, Rahayu; Lestari, Diyah Erlina; Sriyono, Sriyono; Sunaryo, Geni Rina
SIGMA EPSILON - Buletin Ilmiah Teknologi Keselamatan Reaktor Nuklir Vol 20, No 2 (2016): November 2016
Publisher : Badan Tenaga Nuklir Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (544.992 KB) | DOI: 10.17146/sigma.2016.20.2.3513

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The G.A Siwabessy reactor (RSG-GAS) located in the Puspiptek area uses water as a coolant. The water as a coolant will contact directly with the component or structure of the reactor, that a chemical reac- tion between water and those components might cause the possibility of corrosion process. Therefore, cooling water quality will determine the integrity of reactor components or structures. The research described in this paper was conducted in order to monitor the quality of secondary cooling water, so that the water quality specifications is maintained and the reactor can be safely operated. One way to monitor the cooling water quality is by performing analysis into the secondary cooling water and raw water on June 6, 2016. The methodology used was by analysing the pH value using a pH-meter, conductivity value using Conductivity-meter, water hardness analysis, and analysis for some chemical elements such as Cl-, SO42-, Fe, P using calibrated Spectrophotometer DR / 2400. Corrosion rate of the carbon-steel as the piping material of secondary cooling system under environmental corrosion condition was also analyzed using the Potentiostat. From those performed analysis, the overall measured values are still below the standard values as required in the RSG-GAS safety analysis report document, meaning that the water quality management of the secondary coo- ling system has been well performed so far. 
THE STRATEGY TO SUPPORT HTGR FUELS FOR THE 10 MW INDONESIA'S EXPERIMENTAL POWER REACTOR (RDE) Taryo, Taswanda; Ridwan, Ridwan; Sunaryo, Geni Rina; Rachmawati, Meniek
Urania Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol 24, No 1 (2018): Februari, 2018
Publisher : Pusat Teknologi Bahan Bakar Nuklir

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (660.685 KB) | DOI: 10.17146/urania.2018.24.1.3729

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The Indonesia?s 10 MW experimental power reactor (RDE) is developed based on high temperature gas-cooled reactor (HTGR) and the program of the RDE was firstly introduced to the Agency for National Development Planning (BAPPENAS) at the beginning of 2014. The RDE program is expected to have positive impacts on community prosperity, self-reliance and sovereignty of Indonesia. The availability of RDE will be able to accelerate advanced nuclear power technology development and hence elevate Indonesia to be the nuclear champion in the ASEAN region. The RDE is expected to be operable in 2022/2023. In terms of fuel supply for the reactor, the first batch of RDE fuel will be inclusive in the RDE engineering, procurement and construction (RDE-EPC) contract for the assurance of the RDE reactor operation from 2023 to 2027. Consideration of RDE fuel plant construction is important as RDE can be the basis for the development of reactors of similar type with small-medium power (25 MWe?200/300 MWe), which are preferable for eastern part of Indonesia. To study the feasibility of the construction of RDE fuel plant, current state of the art of the R&D on HTGR fuel in some advanced countries such as European countries, the United States, South Africa and Japan will be discussed and overviewed to draw a conclusion about the prospective countries for supporting the fuel for long-term RDE operation. The strategy and roadmap for the preparation of the RDE fuel plant construction with the involvement of national stakeholders have been developed. The best possible vendor country to support HTGR fuel for long-term operation is finally accomplished. In the end, this paper can be assigned as a reference for the planning and construction of HTGR RDE fuel fabrication plant in Indonesia.Keywords: RDE, Indonesia, HTGR, fuel, strategy.