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Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir
Published by BRIN Publishing
ISSN : 08524777     EISSN : 25280473     DOI : https://doi.org/10.55981/urania
Core Subject :
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir is a scientific publication platform that presents experimental research and development studies on the nuclear fuel cycle, covering radiation physics and chemistry, reactor engineering, nuclear safety and security, nuclear forensics, and radioactive waste management
Arjuna Subject : -
Articles 183 Documents
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PENGARUH KONDISI PENYIMPANAN TERHADAP KETAHANAN KOROSI DRUM BAJA KARBON WADAH LIMBAH RADIOAKTIF Dwi Luhur Ibnu Saputra; Aisyah Aisyah; Risdiyana Setiawan; Pungky Ayu Artiani; Kuat Heriyanto; Jaka Rachmadetin
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 29 No. 1 (2023): 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.2023.29.1.6770

Abstract

The Radioactive Waste Management Installation (IPLR) in the Serpong Nuclear Area is Indonesia’s only radioactive waste management facility. This facility has treated various types of waste using carbon steel containers, including low and medium-contamination solid waste. Radioactive waste containers must have good corrosion resistance so that the waste container can survive a long storage period. Research on the corrosion of carbon steel radioactive waste containers has been carried out. This study aims to obtain data on the corrosion rate of welded materials radioactive waste container and their parent metals under disposal conditions. In the disposal facility, there is a potential for water seeps in so that the interaction with the waste container may occurred. This condition will increase the corrosion rate. In this study measured the corrosion rate in various corrosion media, namely SP4 water, demineralization water, cemented solution, and bentonite solution. The results showed that the corrosion rate of the welded container was higher than the corrosion rate of its parent metal in various corrosion media. The highest corrosion rate is SP4 water media, which is 7.86 and 6.78 times compared to demineralization water. Meanwhile, the corrosion rate in demineralization water media, bentonite water, and cement water is not significantly different. The corrosion rate is an essential parameter of radioactive waste containers because waste containers with high corrosion resistance will be able to confine radionuclides for a long period time.
SAFETY INTEGRITY LEVEL ASSESSMENT AT URANIUM EVAPORATOR AND DEPOSITION VESSEL IN NON NUCLEAR REACTOR INSTALLATIONS Putra Oktavianto; Anis Noor Kundari; Ade Saputra; Imam Abdurrosyid; Munisatun Sholikhah; Andri Saputra
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 29 No. 1 (2023): 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.2023.29.1.6986

Abstract

The operation of non-reactor nuclear installations that use nuclear material in the process must be ensured safely during the process. One of the assessments of the safety level of the protection system that has been owned by the installation is using the Safety Integrity Level (SIL) which assesses the safety level of the protection system based on the value of the risk reduction factor that the protection system can achieve. The uranium evaporator and deposition vessel at the Experimental Fuel Element Installation (EFEI) is one of the installations that uses nuclear material in the process so it is necessary to assess the SIL of these 2 vessels. The piping and instrumentation diagram (P&ID) is used to determine the SIL value and Safety Instrumented System (SIS) component are installed in evaporator and deposition vessel. Maintenance data and OREDA (Offshore Reliability Data) are used to determine the failure rate. After knowing and determining the installed SIS components, then determining the architecture vote of the Safety Instrumented Function (SIF) based on the P&ID diagram, so that it can be known that the installed SIS uses vote 1oo1, 1oo2, 1oo3, or the appropriate vote. The installed SIF architecture vote will determine the equation used to calculate the Probability Failure on Demand (PFD). The total PFD obtained is adjusted to the SIL table to find out what SIL level the installed protection system is at. The value of the safety level of the protection system with SIL assessment in the evaporator vessel obtained 2 protection systems with SIL values of level 2 all and in the deposition vessel obtained 2 protection systems with SIL values of level 1 and level 2. The SIL value in the evaporator and deposition vessel analyzed has not reached level 3 or 4, so it is necessary to add SIF to the SIS protection system to increase the SIL value until the SIL value is obtained between level 3 or 4 because safety in the operation of non-reactor nuclear installations is absolute.
ANALISIS KRITIKALITAS PROSES HYDRIDING SERTA PENYIMPANAN PADUAN UMo DAN UZr SEBAGAI KANDIDAT BAHAN BAKAR REAKTOR RISET MENGGUNAKAN OPENMC Saga Octadamailah; Anissa Isnaini; Rohmad Sigit
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 29 No. 1 (2023): 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.2023.29.1.6976

Abstract

Current research on research reactor fuel is focused at increasing uranium density to produce higher neutron flux in order to increase radioisotope production as well as efficiency and accuracy in various material tests. Research and developments of nuclear fuel in Indonesia in order to increase uranium density is the development of UMo and UZr. One of the processes for making UMo and UZr powder is through a hydriding-dehydriding process which involves enriched uranium, and high pressure hydrogen gas. This combination presents the potential for criticality accidents that must be anticipated. Criticality calculations were carried out using OpenMC software, which is a nuclear physics simulation that uses the Monte Carlo method. In this simulation, two types of alloys are used, namely U7Mo and U6Zr. Simulations were carried out using 10,000 particles and 44 batches as parameter settings. From the simulation results, the keff value for the U7Mo and U6Zr alloy are 0.0297 and 0.0314, respectively. Both values are far below the critical value, indicating that the potential for criticality accidents is very low. In the storage simulation of hydriding products, there is the potential for criticality accidents if the storage bottles are placed close together without any distance. To anticipate this risk, each bottle must be given a safe distance and there needs to be a limit on the permitted mass of uranium per unit volume.
POTENSI LINGKUNGAN KAWASAN NUKLIR SERPONG DAN EVALUASI KESELAMATAN UNTUK TAPAK FASILITAS BOREHOLE DISPOSAL LIMBAH SUMBER RADIASI BEKAS Risdiyana Setiawan; Kartini Megasari; Sucipta Sucipta; Andry Setiawan
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 29 No. 1 (2023): 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.2023.29.1.6717

Abstract

Disposal of used radiation sources (DSRS) still faces several obstacles, among others, because DSRS generally have a long half-life, high levels of gamma-photon radiation, difficult options for returning to the country of manufacture, and activity that exceeds the limit for shallow land disposal. For this reason, it is necessary to develop a disposal system for DSRS waste with small-scale facilities that are economical, meet worker, community and environmental safety standards. The answer to this problem is the concept of borehole disposal (BHD). This study focuses on evaluating the environmental potential of the Serpong Nuclear Area (KNS) for the BHD facility site and its safety according to the IAEA standard, namely borehole disposal of disused sealed sources (BOSS). Environmental evaluation of KNS is carried out by characterization and evaluation of geological and non-geological environmental safety for BHD sites according to IAEA standards. In general, the results of the evaluation of environmental parameters against the criteria for the disposal site indicate a suitability. However, there are several parameters that need to be given a technological solution (engineered barrier) so that the site's function as a natural barrier can be fulfilled. Some of these technological solutions include waste packaging (capsules and containers), buffer and backfill materials, borehole linings (casing), basement plug and upper/cover plug installation. The results of the safety assessment using Borehole Disposal Concept (BDS) scoping tools show that the BHD facility concept at the KNS site is included in the green zone which means it is safe, with a maximum total dose value of 7.2e-9 Sv/y in 46,678 y. Using the BOSS standard system with the addition of an engineered barrier, the site location at KNS meets the criteria and has the potential for placement of the DSRS waste borehole disposal facility.
INVESTIGASI MODIFIKASI LUBANG PENGISIAN GAS HELIUM PADA DESAIN END PLUG BATANG UJI BAHAN BAKAR NUKLIR MENGGUNAKAN ANALISIS COMPUTATIONAL FLUID DYNAMICS Hanif Ghufron; Shofwan Bahar; Sugeng Rianto; Praditya Vankabo; Gagad Rahmadi; Odi Buana Hutapea; Triarjo Triarjo; Iwan Setiawan
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 29 No. 1 (2023): 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.2023.29.1.6956

Abstract

A modified design of the helium gas filling hole in the end plug of the PRTF test rod was investigated using CFD simulation method. The simulation is conducted to predict whether helium gas can flow properly into the cladding through the modified design. From this simulation, it is expected that data will be obtained on the effect of changes in the shape and dimensions of the helium gas filling hole. The data will then be used as a guide to determine whether the modified design can function better so that it can replace the original design. The input data required for CFD simulation are the geometry of the model, the type of material used, and the condition parameters for the preparation of the welding process such as pressure and time. Then, the CFD simulation process is carried out in accordance with the conditions that occur in the TIG orbital welding machine. The simulation results show that the modified design is able to flow helium gas into the cladding twice as fast as the original design. In addition, the simulation results also show that the modified design is as stable as the original design in maintaining helium gas pressure at 1.5 bar.
EFEKTIVITAS DESORPSI CESIUM OLEH SURFAKTAN KATIONIK BERDASARKAN STRUKTUR HIDROFILIKNYA PADA MONTMORILLONIT K10 Herry Wijayanto; Satoru Nakashima
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 29 No. 1 (2023): 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.2023.29.1.6962

Abstract

This study discusses the efficiency of cesium (Cs) desorption from montmorillonite K10 samples (MMTK10) previously contaminated with Cs (MMTK10/Cs) using cationic surfactant desorbents, benzyldodecyldimethylammonium bromide (BDAB) and dodecyltrimethylammonium bromide (DTAB), as well as HCl as a desorption control. The aim of this research is to explore the effects of the different hydrophilic structures of cationic surfactants on Cs desorption from MMTK10, where BDAB has a larger hydrophilic structure compared to DTAB. The results show that cationic surfactants, particularly at high concentrations of 20-50 mM, exhibit higher desorption efficiency compared to HCl, with efficiencies reaching 80-90%. This mechanism is associated with the ability of cationic surfactants to form micelles when their concentration exceeds the critical micelle concentration (CMC) or at high concentrations, which facilitates ion exchange with Cs on the surface of MMTK10/Cs. There is no significant difference in Cs desorption efficiency between BDAB and DTAB, even though the hydrophilic structure of BDAB is larger. This contrasts with previous studies on other montmorillonites, where a larger hydrophilic surfactant structure led to greater opening of the montmorillonite interlayer due to the entry of the larger surfactant, resulting in greater Cs desorption. The lack of a difference in Cs desorption efficiency between the two surfactants in this study is due to the specific characteristics of MMTK10 clay itself, which does not have the ability to expand or enlarge the interlayer due to heat treatment during its production before being sold in the market. Additionally, the similarity in the length of the hydrophobic tail chains between BDAB and DTAB contributes to the identical Cs desorption results from both surfactants. In conclusion, while the nature and structure of cationic surfactants influence the Cs desorption process, the specific characteristics of the used montmorillonite or clay also play a crucial role in this desorption mechanism.
IMAGE ENHANCEMENT CITRA DIGITAL SHORT PIN PWR FUEL BERISI PELET UO2 ALAM PASCA UJI RADIOGRAFI SINAR-X MENGGUNAKAN PROGRAM IMAGEJ Refa Artika; Rohmad Sigit; Maman Kartaman Ajiriyanto; Tri Yulianto
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 28 No. 1 (2022): 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.2022.28.1.6617

Abstract

Non-destructive test using digital X-ray radiography is one of the important series in performance evaluation of nuclear materials and fuels. Image enhancement or image quality improvement is needed for further analysis of digital X-ray radiography images which tend to have lower resolution. Image enhancement process on digital x-ray radiographic images needs to be carried out to provide images that are easier to interpret so that the resulting data can be used as a basis for evaluating performance after irradiation. The tests were performed at a distance of 40 cm against the detector at voltage and current of 120 kV and 1000 μA. The image enhancement was later performed using ImageJ, which involves the process of aligning the image contrast, edge enhancement, processing histogram value, and adjusting image colors. The resulted data were costumized according to the information to be obtained from the digital X-ray radiography test such as dimensional data, boundaries between pellet, and histogram ilustrating material or thickness differences. Further, this information can be used as data input in mapping the concept of enhancement of image of short pin PWR fuel contains UO2 pellets. Output image produced from the enhancement process not only has a better resolution and quality to facilitate the process of analysis and interpretation but also can be used as a standard method to perform image enhancement of digital X-ray radiography images of short pin PWR fuel contains UO2 pellets after irradiation.
PENGARUH PENAMBAHAN LARUTAN H3BO3 DAN LiOH TERHADAP PERILAKU KOROSI MATERIAL KELONGSONG ZIRCALOY-2 DALAM MEDIA AIR BEBAS MINERAL Fajar Al Afghani; Erwan Hermawan; Anditania Sari Dwi Putri; Ely Nurlaily
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 28 No. 1 (2022): 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.2022.28.1.6609

Abstract

The primary coolant of the PHWR type reactor is controlled chemically by the addition of boric acid and LiOH to anticipate corrosion of the cladding material of Zr-2. This study aimed to analyze the effect of H3BO3 and LiOH solutions on the type and rate of of zircaloy-2 (Zr-2) corrosion in demineralized water. The corrosion test was carried out at room temperature to eliminate the influence of high temperature and pressure on the electrochemical process. In this study, the corrosion rate of Zr-2 was observed in a demineralized water solution with varying concentrations of H3BO3 and LiOH. The observation was done using Tafel method, where the addition of H3BO3 at concentrations of 10 ppm, 100 ppm, 500 ppm,1000 ppm and 2000 ppm resulted in corrosion rate of 17.29 x 10-3 mpy, 18. 51 x 10-3 mpy,20.82 x 10-3 mpy, 22.71 x 10-3 mpy and 23.29 x 10-3 mpy, respectively. The addition of 2000 ppm of H3BO3 and LiOH with varied concentrations of 1 ppm, 2 ppm, 3 ppm, 4 ppm results in decreasing of corrosion rate sequentially, namely 22.71 x 10-3 mpy, 21.88 x 10-3 mpy, 21.41 x 10-3 mpy, 21.39 x 10-3 mpy, whereas the addition of 5 ppm LiOH causes the corrosion rate to increase to 21.45 x 10-3 mpy. It can be concluded that the addition of LiOH and H3BO3 affects the corrosion rate of zircaloy-2 in demineralized water. Higher concentration of H3BO3 causes the corrosion rate to increase meanwhile, the addition of LiOH causes the corrosion rate to decrease to a concentration of 4 ppm. However, the addition of LiOH at a concentration of 5 ppm causes the corrosion rate to increase.
NEUTRONIC CALCULATION FOR PWR MOX FUEL PIN CELLS WITH WIMSD-5B CODE Santo Paulus Rajagukguk; Syaiful Bakhri; Tukiran Surbakti
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 28 No. 1 (2022): 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.2022.28.1.6610

Abstract

The WIMSD-5B thermal reactor lattice cell code is used in many laboratories for research reactor calculations and power reactors. The program uses the Wigner-Seitz approximation for cell pin calculations. The approximation has been widely applied to the pin of UO2 cells and has shown good results in previous studies but can produce incorrect results when used for pin cells in MOX fuels. This paper investigates the use of the WIMS-5B code to calculate the neutron multiplication factor and depletion for MOX fuel pin cells. Calculations were performed using the WIMSD-5B code updated with the ENDF-BVIII.0 library. The outer scattering boundary condition was used to overcome the effect of the Wigner-Seitz approach on the lack of MOX fuel. Results of this study indicates that most of the results obtained using ENDF-BVIII.0 are better than ENDF-BVII.1, and this is in line with expectations. The difference in the maximum k-inf value obtained from this library occurs in the fuel that has the greatest enrichment. On the other hand, the addition of the outer scattering limit improves the results obtained using ENDF-BVIII.0, causing a slight improvement for other libraries. This shows that by using appropriate libraries and the addition of the scattering outer limit, the Wigner Seitz approximation for the MOX pin cell pins in WIMS-D5 can yield quite accurate results.
EFFECTS OF FUEL DENSITY ON REACTIVITY COEFFICIENTS AND KINETIC PARAMETERS OF PEBBLE BED REACTOR Suwoto; R. Andika Putra Dwijayanto; Wahid Luthfi; Zuhair
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 28 No. 1 (2022): 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.2022.28.1.6625

Abstract

Few decades ago, a large number of nuclear reactors were designed to use HEU as the fuel. But the use of HEU is being discouraged since it can be used as a nuclear explosive material which makes its proliferation potential. Most of the HEU-fueled nuclear reactors in the world are either closed or converted into other types that have LEU fuel with 235U enrichment below 20%. To extend lifetime, LEU fuel with high density is developed. The change in fuel density from low to high will also change core neutronics and thermal hydraulics of the reactor, and as a result, it affects the transient response of the reactor. This paper studies the effects of fuel density on reactivity coefficients and kinetic parameters of pebble bed reactor through several calculations with MCNP6 code combine with ENDF/B-VII.1 continuous energy cross-section nuclear data library. The overall calculation results show that the Doppler temperature coefficient (DTC) increases with increasing fuel density, but the moderator temperature coefficient (MTC) decreases due to hardening of neutron spectrum. Kinetic parameters such as effective delayed neutron fraction (βeff), prompt neutron lifetime(ℓ) and neutron generation time (Ʌ) which significantly reduced with increasing fuel density will strongly affect the reactor control and safety. The results of this work conclude that the selection of 9–15 g/cm3 fuel density should be considered carefully given that its effect on reactor controllability.

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