Atom Indonesia (E-Journal)
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Synthesis of Polyvinyl Alcohol (PVA)-Gelatin Hydrogel from White Snapper (Lates calcarifer, Bloch) with Gamma Irradiation and Its Characterizations
Full text linkThe application of nuclear technology in the health sector is increasing. One example is the use of irradiation in production of wound dressings. Research activities have been conducted to study whether polyvinyl alcohol (PVA)-gelatin-based hydrogel from white snapper scales can be processed using gamma irradiation into wound dressings. A series of PVA (10 %) solutions containing gelatin in various concentrations (0-4 %) were treated with three freeze-thaw cycles and then irradiated at doses of 10 and 20 kGy. They were subsequently characterized using Fourier transform infrared (FTIR) spectroscopy and scanning electron microscope (SEM). Gel fraction, water absorption, and percentage of hydrogel water evaporation rate were tested gravimetrically, while the elongation at break and the tensile strength of the hydrogels were tested with a universal testing meter. The evaluation showed that the hydrogel gel fraction decreased with increasing gelatin concentration from 0 % to 4 % for both irradiation doses (10 and 20 kGy). The rising gelatin concentration demonstrated that increasing gamma radiation dose improved the hydrogel's water absorption, evaporation rate, tensile strength, and elongation at break. PVA-gelatin hydrogel with irregular pore structure was observed from SEM test results. The FTIR measurement results confirmed the formation of crosslinks in the hydrogel matrix. The PVA-gelatin hydrogel produced through gamma irradiation could be used for wound dressings
Uncovering the Distribution Zones of Uranium and Thorium in Bangka Island
Full text linkRadioactive minerals, especially containing uranium and thorium, can be used as a core element of nuclear fuel. Bangka Island is located in The Southeast Asian Tin Belt where it has a large uranium and thorium potential. The purpose of this study is to delineate distribution zone of uranium and thorium in Bangka Island. The study methods consist of radiometric measurement and mapping, petrographic analysis, and mineralogical analysis of pan concentrate samples. Based on radiometric measurement, positive anomaly value of equivalent uranium (eU) is ranging from 5-15 ppm while of equivalent thorium (eTh) is ranging from 45-75 ppm. The result of petrographic analysis from several outcrops of Klabat Granite indicated that there are monazites found in several samples of Mangkol Granite and of Bebuluh Granite. Radioactive mineral indication also can be identified as pleochroic halo within biotite in samples of Pelangas Granite and Menumbing Granite. Based on the result of mineralogical analysis of pan concentrate samples, it was identified that monazites can be found in all samples. Monazites constitute the percentages ranging from 2.82-10.66 %. Zircon also can be identified with percentages ranging from 9.13-76.75 % while ilmenite and magnetite minerals have average percentages of 24.09 % and 5.97 %, respectively. Favorable zones can be delineated in outcrops of Klabat Granite, Ranggam Formation and alluvial deposits in northern, northwestern, northeastern, central, and southeastern parts of Bangka Island. The occurrences of monazites in those lithological units are the main factors of high radioactivity in Bangka Island. Based on petrographic and mineralogical composition, those granite bodies which are correlated with Klabat Granite are mostly associated with ilmenite series with S-Type granitic rocks
Neutronic Parameter Analysis of Plate-Type Fueled TRIGA 2000 Reactor by MCNPX
Full text linkA novel simulation to calculate the neutronic parameters of the TRIGA 2000 reactor using plate-type fuel has been performed. The plate fuel used was produced by the Indonesian Nuclear Industry (PT INUKI) with U3Si2-Al material. Neutronic parameters based on INUKI’s plate-type fuel dimension and the current TRIGA’s configuration were simulated using MCNPX. The simulation was performed by modeling the complete reactor’s configuration on a fresh fuel core state. We obtained the kinetic parameter values from the simulation, i.e., delayed neutron fraction of 8.11×10‑3, a prompt neutron lifetime of 2.0551×10‑4 s, and an average neutron generation time of 1.87×10‑4 s. The excess reactivity of the reactor was 9.02 %Δk/k, while reactivity in the one-stuck-rod state was below ‑0.5 ). The average thermal neutron flux peak occurred at the central irradiation position with the value of 3.0×1013 to 3.1×1013 n/(cm2 s). The reactor has a power peaking factor of 1.379 in the control rod position of 0 % on D3 fuel. The reactor had a negative feedback reactivity coefficient, except for the moderator coefficient. These results suggest that the current configuration of plate-type fuel met the nuclear reactor neutronic safety standards
Optimizing Neutronic and Photonic Performance in Irradiation Systems of Symmetric TRIGA Cores
Full text linkThe BAEC TRIGA MARK-II Research Reactor (BTRR) in Bangladesh has been used for a wide range of purposes, including basic and applied nuclear research and human resource development. Therefore, its core management should be flexible to meet various objectives with different priorities and to deliver the best possible outcome. In this study, neutron and gamma photon flux variation was studied at different radial and axial irradiation systems of the current core (C-0) as well as six symmetric reconfigurations (C-1, C-2, C-3, C-4, C-5, and C-6) of the existing BTRR using the universal MCNP code. While keeping the exact core component and material density, the symmetric reconfigured cores were modeled based on core criticality calculation and excess reactivity in the critical state. Finally, it was observed that the reconfigured core C-1 has the best neutronic and photonic performance at the irradiation systems compared to other reconfigured cores, against the reference core C-0
Assessment of 137Cs in the Environment of Hetauda City, Nepal by In-Situ Gamma Ray Spectrometry
Full text linkA significant amount of 137Cs radioactive fallout have been spread in the atmosphere due to nuclear weapon testing and nuclear reactor disasters. This fallout eventually settles on the Earth's surface, and because 137Cs has a long half-life, it remains in the environment for an extended period. Mapping the distribution of 137Cs is crucial, and this study aims to assess the radioactive deposition of 137Cs in the ground to establish baseline data for its distribution in the environment of Hetauda City, Nepal. Recently, Hetauda City has been designated as the capital city of the Bagmati province. To measure 137Cs deposition, portable (backpack) gamma ray spectrometer was used with a 0.347-liter NaI(Tl) detector. Rapid measurement was carried out while walking at a pace of less than 2 km/h, and the distance between the detector and the ground was maintained at less than 1 m with the detector pointing downward. The surface activity of 137Cs was measured in the range of 0.003 to 2.382 kBq/m2, with an average value of 0.581 ± 0.343 kBq/m2. The spatial variability of 137Cs was found to be smooth in the area, and the mean annual effective dose calculated was 0.379 ± 0.224 µSv. The low dose rates and smooth spatial distribution of 137Cs in the environment indicate no contamination, and the trace amount present could be due to global fallout from weapons testing and nuclear accidents. The results were compared with previously reported values worldwide
Dose Response of Personnel OSL Dosimeter to the Cesium-137 and 80 kVp X-ray Exposure
Full text linkOver the years, several types of dosimeters have been introduced for accurate dose assessment. The OSL dosimeter is one of them. It is used to monitor personnel dose from external exposure. In this paper, dose response of OSL dosimeters in terms of Hp(10) to Cs‑137 gamma and 80 kVp X‑ray radiation will be studied. The dosimeters were irradiated using Cs‑137 gamma and 80 kVp X‑ray to 0.5 mSv, 1 mSv, 3 mSv, 5 mSv, and 10 mSv at a distance of 200 cm, and all of them were subsequently read. Half of the dosimeters that were previously irradiated with a dose of 1 mSv and 5 mSv were read 30 times. The other half of the dosimeters were re-read on day 30 and day 60 from the initial reading. The study shows that relations between measured dose and exposure dose for Cs‑137 gamma and 80 kVp X‑ray irradiation are linear with correlation coefficients (R2) of 0.9997 and 0.9987, respectively. When the OSL dosimeters were read repeatedly, a dose reduction for each reading occured by 0.4 % and 0.5 % on Cs‑137 gamma and 80 kVp X‑ray, respectively. Dose reading on day 60 after Cs‑137 gamma irradiation showed fading of 3.6 % and 2.7 % on OSL dosimeter exposed to 1 mSv and 5 mSv, respectively, whereas fading effect on 80 kVp X‑ray irradiation showed values of 5.9 % and 8.8 % for the two doses
Neutronic Evaluation of Using a Thorium Sulfate Solution in an Aqueous Homogeneous Reactor
Full text linkRadioisotope 99Mo is one of the most essential radioisotopes in nuclear medicine. Its production in an Aqueous Homogeneous Reactor (AHR) could be potentially advantageous compared to the traditional technology, based on target irradiation in a heterogeneous reactor. An AHR conceptual design using low-enriched uranium for the production of 99Mo has been studied in depth. So far, the possibility of replacing uranium with a non-uranium fuel, specifically a mixture of 232Th and 233U, has not been evaluated in the conceptual design. Therefore, the studies conducted in this article aim to evaluate the neutronic behavior of the AHR conceptual design using thorium sulfate solution. Here, the 232Th-233U composition to guarantee ten years of operation without refueling, conversion ratio, medical isotopes production levels, and reactor kinetic parameters were evaluated, using the computational code MCNP6. It was obtained that 14 % 233U enrichment guarantees the reactor operation for ten years without refueling. The conversion ratio was calculated at 0.14. The calculated 99Mo production in the AHR conceptual design resulted in 24.4 % higher with uranium fuel than with thorium fuel
A Study on Radiation Hazard of Granite and Marble Widely Used in Jordan Using Gamma Ray Spectrometer
Full text linkGranite and marble are widely used in building construction, so possible radioactive nuclides inside them may contribute to the exposure dose to human health. The purpose of this study was to investigate the natural radioactivity concentration and assess the radiological risk limits and health care. The samples of marble and granite were pulverized into small, fine, smooth pieces and counted with the GAMMA-X (GMX) spectrometer to measure the radioactivity concentrations of 238U, 232Th, and 40K. The radiological dose, internal and external hazards, and radium equivalent activity were calculated with a standard formula. The results showed that the radioactive concentrations of 238U, 232Th, and 40K in granite were higher than those in marble. The external hazard for granite samples was below unity, while its internal hazard exceeded unity. The radium equivalent activity did not exceed the critical legal level of 370 Bq/kg as a safe level. For marble, the external and internal hazards and radium equivalent activities showed good agreement with the safe construction level. Its external and internal hazards were less than unity, whereas the radium equivalent activities were less than the critical legal level
Assessment of TMSR-500 Shutdown Capability
Full text linkThe molten salt reactor (MSR) is a generation IV reactor with liquid fuel having nearly zero excess reactivity. Due to the very low excess reactivity, it requires a small number of control rods worth to shut down the reactor. However, as it operates at high temperatures, the core reactivity increases as the fuel temperature cools down during shutdown. In such a case, the control rods might not be able to keep the reactor at a subcritical state, and consequently, the fuel must be removed from the core for long-term shutdown into a fuel drain tank (FDT) below the core. This paper is intended to assess the shutdown capability of the first active shutdown system and fuel drain tank of ThorCon MSR by doing neutronic calculations with MCNP6. The results indicated that the control rods having reactivity worth ‑1.699 %dk/k are unable to maintain the core at a subcritical state as the core excess reactivity increases to +7.760 %dk/k when the fuel reaches room temperature. Therefore, the fuel must be drained to FDT to be cooled down and kept subcritical. Evaluation for various cases of FDT produced the highest multiplication factor of 0.57008 ± 0.00004 at the most conservative condition. The multiplication factor is well below the critical state of 1.0. The evaluations suggest that soon after the control rods shut the reactor down, the fuel has to be drained to FDT to maintain shutdown condition and dissipate the decay heat