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All Journal Marine Fisheries: Jurnal Teknologi dan Manajemen Perikanan Laut Jurnal Teknologi Reaktor Nuklir Tri Dasa Mega Ganendra: Majalah IPTEK Nuklir Urania Jurnal Ilmiah Daur Bahan Bakar Nuklir ETERNAL: English Teaching Journal Jurnal Sains Teknologi dalam Pemberdayaan Masyarakat Jurnal Bluefin Fisheries Jurnal Airaha Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir
Yuli Purwanto
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Agriculture, Biological Sciences & Forestry Humanities Astronomy Biochemistry, Genetics & Molecular Biology Chemistry Computer Science & IT Control & Systems Engineering Education Energy Engineering Environmental Science Industrial & Manufacturing Engineering Languange, Linguistic, Communication & Media Materials Science & Nanotechnology Mechanical Engineering Social Sciences Other

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DESIGN OF DRY CASK STORAGE FOR SERPONG MULTI PURPOSE REACTOR SPENT NUCLEAR FUEL Dyah Sulistyani Rahayu; Yuli Purwanto; Zainus Salimin
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 24 No. 1 (2018): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/urania.2018.24.1.3694

Abstract

The spent nuclear fuel (SNF) from Serpong Multipurpose Reactor, after 100 days storing in the reactor pond, is transferred to water pool interim storage for spent fuel (ISFSF). At present there are a remaining of 245 elements of SNF on the ISSF,198 element of which have been re-exported to the USA. The dry-cask storage allows the SNF, which has already been cooled in the ISSF, to lower its radiation exposure and heat decay at a very low level. Design of the dry cask storage for SNF has been done. Dual purpose of unventilated vertical dry cask was selected among other choices of metal cask, horizontal concrete modules, and modular vaults by taking into account of technical and economical advantages. The designed structure of cask consists of SNF rack canister, inner steel liner, concrete shielding of cask, and outer steel liner. To avoid bimetallic corrosion, the construction material for canister and inner steel liner follows the same material construction of fuel cladding, i.e. the alloy of AlMg2. The construction material of outer steel liner is copper to facilitate the heat transfer from the cask to the atmosphere. The total decay heat is transferred from SNF elements bundle to the atmosphere by a serial of heat transfer resistance for canister wall, inner steel liner, concrete shielding, and outer steel liner respectively. The rack canister optimum capacity of 34 fuel elements was designed by geometric similarity method based on SNF position arrangement of 7 x 6 triangular pitch array of fuel elements for prohibiting criticality by spontaneous neutron. The SNF elements are stored vertically on the rack canister. The thickness of concrete wall shielding was calculated by trial and error to give air temperature of 30 oC and radiation dose on the wall surface of outer liner of 200 mrem/h. The SNF elements bundles originate from the existing racks of wet storage, i.e. rack canister no 3, 8 and 10. The value of I0 from the rack no 3, 8 and 10 are 434.307; 446.344; and 442.375 mrem/h respectively. The total heat decay from rack canister no 3,8 and 10 are 179.640; 335.2; and 298.551 watts. The result of the trial-and-error calculation indicates that the rack canister no 3, 8 and 10 need the thickness of concrete shielding of 0.1912, 0.1954 and 0.1940 m respectively.
EVALUATION OF NEUTRON SHIELDING PERFORMANCE OF CD-SS 316L AS A CANDIDATE ALLOY FOR DRY CASK OF RESEARCH REACTOR SPENT FUEL Kuat Heriyanto; Usman Sudjadi; Jaka Rachmadetin; Yuli Purwanto; Pungky Ayu Artiani; Siti Chotijah
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 27 No. 2 (2021): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/urania.2021.27.2.6426

Abstract

Development of dry casks is necessary to support national strategy for the management of spent fuels. One of the requirements for the dry cask is shielding performance for neutrons emitted by the spent fuels. The objectives of this study were to determine the emitted neutrons by the spent fuels generated from GAS research reactor and to evaluate the neutron shielding performance of Cd-SS316L alloy as a candidate material to be used for dry cask for the spent fuels. The determination of the emitted neutrons was carried out using Origen 2.1 software while the evaluation of the neutron shielding performance was done using MCNP5. The result shows that the emitted neutrons by a spent fuel after 5 years discharged from GAS research reactor are 2.81×103 and 3.32×106 n/s for reactor core power of 15 and 30 MW, respectively. Addition of Cd improves the neutron shielding performance of SS 316L. It can be concluded that evaluation of neutron shielding performance of Cd-SS316L can be done using Origen 2.1 software for the neutron emission determination and MNCP 5 software for the neutron shielding performance by simulation. This study shows that the Cd-SS 316L alloy is worth further study to develop the dry cask design for the GAS research reactor
PENGARUH IRADIASI GAMMA TERHADAP KARAKTERISTIK POLIMER POLIESTER TAK JENUH UNTUK APLIKASI SOLIDIFIKASI LIMBAH RADIOAKTIF Risdiyana Setiawan; Purwantiningsih Sugita; Muhammad Farid; Ratiko; Jaka Rachmadetin; Kuat Heriyanto; Yuli Purwanto; Dwi Luhur Ibnu Saputra; Andry Setiawan
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 30 No. 2 (2024): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.55981/urania.2024.9131

Abstract

The first effort in radioactive waste management is through radioactive waste characterization by identifying nuclides and measuring radioactivity. Characterization of radioactive waste is helpful in waste management consisting of waste containment, waste decomposition, temporary waste storage, and disposal operations. The use of polyester polymers in radioactive waste solidification is one of the important solutions in safe and sustainable waste management. These polymers are chosen for their good mechanical and chemical properties, such as resistance to degradation and long-term stability. This study aims to characterize and analyze the effect of gamma irradiation on the mechanical and chemical properties of unsaturated polyester polymers used in radioactive waste solidification. Irradiation was carried out with dose variations of 50, 100, 200, and 300 kGy. The results showed that at low to medium doses (50-200 kGy), there was an increase in compressive strength and density due to the formation of crosslinks between polymer chains. However, at higher irradiation doses (300 kGy), material degradation occurred leading to a decrease in strength. Raman spectrophotometry and SEM showed changes in chemical structure and morphology due to irradiation, especially in terms of changes in functional groups and structural density. The results of this study indicate that irradiation can strengthen polymers at certain doses, but dose limits must be considered to prevent material degradation.
PENGOLAHAN LIMBAH RADIOAKTIF CAIR AKTIVITAS TINGGI MELALUI VITRIFIKASI GELAS BOROSILIKAT Wati; Jaka Rachmadetin; Yuli Purwanto; Dwi Luhur Ibnu Saputra; Ratiko; Aisyah; Pungky Ayu Artiani
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 30 No. 2 (2024): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.55981/urania.2024.9144

Abstract

High-level liquid radioactive waste in Indonesia is presently generated from the Mo99 radioisotope production process. Until present, the treatment of the waste has not been carried out and it is still stored in the hot cell awaiting for the establishment of the methods, technology and processing facilities. Several countries have implemented a vitrification process to process LRCAT. Several types of glass have been developed for LRCAT vitrification, namely alumina silicate glass, phosphate glass, borosilicate glass and ceramic glass. Borosilicate glass is the most ideal type of glass to use. This is because the process of making borosilicate glass is simpler and the results of vitrification of borosilicate waste-glass have chemical resistance and physical resistance as well as high waste content. The objective of this research is to determine the composition of borosilicate glass-forming materials through a vitrification process suitable for imobilation of high-level waste from 99Mo production that meets the requirements of the final waste form. Several requirements in choosing a matrix for LRCAT vitrification are: easy and practical manufacturing process, high waste content, high chemical resistance, high radiation stability, high heat stability, good physical integrity and long-term stability. The vitrification process uses borosilicate glass forming materials with two different compositions. The vitrification process was carried out at a temperature of 1150 ºC for 2.5 hours then cooled to room temperature. Glass characterization is carried out by leaching test and X-ray diffractometer (XRD). Waste glass with GF-1 glass forming material with various waste contents cannot be melted at a temperature of 1150 ºC. The vitrification process was unsuccessful and no melting of the waste-glass occurred. The glass forming material GF-1 cannot be used for LRCAT vitrification and no further research was carried out. This is related to its use on an industrial scale, if the melting temperature is higher than 1150 ºC, then the refractory corrosion rate is very high so the melter life is shorter. This will generate more solid radioactive waste (secondary waste). In addition, high melting temperature will produce more gas so that gas handling is more complex. Meanwhile, glass material with GF-2 glass forming material can be melted at a temperature of 1150 ºC. GF-1 waste-glass were not leached test because the vitrification process was unsuccessful and no melting of the waste-glass occurred. The leaching test results of all GF-2 glass-waste compositions that were made ranged from 0.79 x 10-4 gram.cm-2.day-1 to 2.25 x 10-4 gram.cm-2.day-1. The leach rate requirement for PNC Japan standard borosilicate glass is 2.30 x 10-5 gram.cm-2.day-1 and based on Pacific Northwest National Laboratory (PNNL) is 10-4 – 10-7 gram.cm-2.day-1. Based on the results of chemical resistance characterization (leachate test) and physical integrity, the composition of GF-2 glass-waste with a waste content of 20% by weight provides the most optimal characteristics
Co-Authors Aisyah Aisyah Aisyah Aisyah Aisyah Andry Setiawan Argaditia Mawadati Budhi H. Iskandar Budy Wiryawan Dwi Luhur Ibnu Saputra Dyah Sulistyani Rahayu Dzakiya, Nurul Emy Setyaningsih Eska Almuntaha Jaka Rachmadetin Johnny H. Tumiwa Karyanto Karyanto Kuat Heriyanto Kuat Heriyanto Malik Musthofa Manohas, Jul Mirawaty Mirawaty Moh. Cecep Cepi Hikmat Mohammad Imron Mohammad Zaini Muhammad Farid Nayandra Dias Fadilla Nidia Lestari P. Ayu Artiani Pratiwi Ayuningtyas Pungky Ayu Artiani Pungky Ayu Artiani Purnawan Purnawan Purwantiningsih Sugita Ratiko Ratiko Ratiko Risdiyana Setiawan Siti Chotijah Suparni Setyowati Rahayu Suwanto Raharjo Untung Sugiharto Usman Sudjadi Wati Wati Wati Zainus Salimin