Turbin angin sumbu vertikal dengan variasi jumlah sudu dan sistem buka tutup sirip adalah suatu alat konversi energi yang mengubah energi gerak dari sudu turbin menjadi energi listrik. Tujuan dari ...tugas akhir ini adalah untuk membuat turbin angin sumbu vertikal tipe sudu bersirip dengan jumlah variasi jumlah sudu dan menguji kinerja turbin angin tersebut. Pada penelitian ini menggunakan metodelogi penelitian kuantitatif dengan cara mengumpulkan data. Variabel bebas pada penelitian ini adalah jumlah sudu 2, 3, 4, 5,6 dan kecepatan angin 7 m/s, 8 m/s, 9 m/s, 10 m/s, 11 m/s, dan 12 m/s. Variabel terikat pada penelitian ini adalah unjuk kerja turbin angin sumbu bersirip yaitu daya kinetik, daya generator dan efisiensi. Sedangkan variabel terkontrolnya adalah jumlah sudu pada turbin angin tipe sudu bersirip sebanyak 6 buah dengan tinggi 600 mm dan lebar 300 mm. Hasil penelitian ini menunjukkan bahwa pengujian menggunakan jumlah sudu yang berbeda-beda mempengaruhi unjuk kerja turbin angin tipe bersirip. Efisiensi sistem tertinggi diperoleh pada jumlah sudu 3 sebesar 1,409 % pada kecepatan 11 m/s.Daya tertinggi pada turbin angin diperoleh dengan jumlah sudu 6 sebesar 6,6 watt pada kecepatan angin 12 m/s dengan putaran generator 202.92 rpm pada beban 30 watt. Berdasarkan analisa, efisiensi maksimal turbin angin hanya dapat dicapai jika intensitas angin konstan sehingga menghasilkan tegangan dan arus yang besar. Kata kunci : Turbin ,Sudu, Sirip, Daya, Efisiensi
Abstract
Several studies related to simplifying the modeling of pebble bed High-Temperature Reactor core (HTR) has been developed before. From some calculation on several MCNP models with a fueled ...pebble to dummy ratio 57:43, using a combination of several types of TRISO (TRi-structural ISOtropic particle fuel) unit and Pebble unit is modeled to achieve its first criticality. In this paper, some MCNP model that uses 27000 pebbles with a 57:43 ratio and 100% fueled pebble is created to be used on burnup calculation and to compare its k-eff and nuclide inventory. From this burnup calculation, it could be seen that SC (Simple Cubic) TRISO unit has faster calculation time followed by the HCP (Hexagonal Close Packed) TRISO unit and then the FCC (Face-Centered Cubic) TRISO unit. The BCC (Body-Centered Cubic) pebble unit had some consistent deviation from another pebble unit, and it still needs more study to know more about the reason behind it. It could be seen that if there are some dummy pebbles inside the reactor, then the deviation would be higher than if there is just fueled pebble inside the reactor. On the 57:43 ratio, the absolute average deviation of k-eff on burnup calculation is lower than 2% and 10% for nuclide inventory (mass). On 100% fueled pebble, it’s below 0.15% on k-eff absolute deviation and below 8% on nuclide inventory deviation.
Specific energy demand of Indonesia with its distributed energy demand required a special energy supply solution. The energy supply solution should be able to fulfil the electric demand of huge and ...growing population which spread in many different islands also to fulfil the energy demand to process abundances yet distributed natural sources. This paper proposes a pebble bed high temperature reactor as an energy supply in the form electric and/or cogeneration application. A parametric survey of the core geometry of a 150MWt Pebble Bed Reactor called PeLUIt is performed, follow by an equilibrium analysis of the optimized design. An accident analysis of depressurized loss of forced cooling (DLOFC) is performed to show the strong passive safety features of the design. Finally the heat-mass balance of the plant is also presented.
The inherent safety feature of a pebble-bed reactor can be observed from its
kinetic parameters. Proper modeling for calculating the reactor kinetic is
also a concern for safe operation during normal ...and transient conditions.
This study is intended to investigate the kinetic parameters
characteristics of a pebble bed reactor using HTR-Proteus. A series of
calculations were conducted using MCNP6 code and ENDF/B-VII library. The
calculation results show that the negative value on core temperature
reactivity is affected dominantly by the Doppler broadening effect. Prompt
neutron lifetime l and mean generation time ? are slightly changed due to an
increase in fuel temperature, moderator, and reflector that changed the
neutron moderation and absorption over this part of the reactor. For (Th,
U)O2, UO2, and PuO2 cores, the effective delayed neutron fraction
?eff values are more influenced by 233U, 235U, and 239Pu, respectively. In
terms of stability during reactivity insertion, the UO 2 core is more stable
and easier to control because its ?eff value is the largest compared to (Th,U)O2
and PuO 2 cores. It can be concluded that temperature must
be controlled because it does not only affect the reactivity but also
kinetic parameters as part of developing inherent safety features on the
pebble-bed reactor.
Radiation damage in core structure of the Indonesian RGS GAS multi purpose reactor resulting from the reaction of fast and thermal neutrons with core material structure was investigated for the first ...time after almost 30 years in operation. The aim is to analyze the degradation level of the critical components of the RSG GAS reactor so that the remaining life of its component can be estimated. Evaluation results of critical components remaining life will be used as data ccompleteness for submission of reactor operating permit extension. Material damage analysis due to neutron radiation is performed for the core structure components made of AlMg3 material and bolts reinforcement of core structure made of SUS304. Material damage evaluation was done on Al and Fe as base metal of AlMg3 and SUS304, respectively. Neutron fluences are evaluated based on the assumption that neutron flux calculations of U3Si8-Al equilibrium core which is operated on power rated of 15 MW. Calculation result using SRAC2006 code of CITATION module shows the maximum total neutron flux and flux >0.1 MeV are 2.537E+14 n/cm2/s and 3.376E+13 n/cm2/s, respectively. It was located at CIP core center close to the fuel element. After operating up to the end of #89 core formation, the total neutron fluence and fluence >0.1 MeV were achieved 9.063E+22 and 1.269E+22 n/cm2, respectively. Those are related to material damage of Al and Fe as much as 17.91 and 10.06 dpa, respectively. Referring to the life time of Al-1100 material irradiated in the neutron field with thermal flux/total flux=1.7 which capable of accepting material damage up to 250 dpa, it was concluded that RSG GAS reactor core structure underwent 7.16% of its operating life span. It means that core structure of RSG GAS reactor is still capable to receive the total neutron fluence of 9.637E+22 n/cm2 or fluence >0.1 MeV of 5.672E+22 n/cm2.
Criticality benchmark experiment at STACY critical facility is important for validation of computation technique and nuclear data library used in design of nuclear fuel cycle criticality safety. This ...paper discusses criticality safety benchmark calculation at STACY facility, which uses uranyl nitrate solution with MCNP-4C Monte Carlo transport code. The continuous energy nuclear data library was utilized in benchmark calculation to complete criticality safety analysis. The MCNP-4C criticality (keff) prediction indicated overestimated results for all configurations except for configuration 131. The biases of calculation with criticality experiment (keff = 1) were under 0.26%. Configuration 140 calculation showed the most precisely agreement with C/E value of 1.0001. From these results, it can be concluded that the capability and reliability of MCNP-4C is constantly high in prediction of criticality accuracy for uranyl nitrate solution at STACY 280T slab core.
Criticality benchmark experiment at STACY critical facility is important for validation of computation technique and nuclear data library used in design of nuclear fuel cycle criticality safety. This ...paper discusses criticality safety benchmark calculation at STACY facility, which uses uranyl nitrate solution with MCNP-4C Monte Carlo transport code. The continuous energy nuclear data library was utilized in benchmark calculation to complete criticality safety analysis. The MCNP-4C criticality (keff) prediction indicated overestimated results for all configurations except for configuration 131. The biases of calculation with criticality experiment (keff = 1) were under 0.26%. Configuration 140 calculation showed the most precisely agreement with C/E value of 1.0001. From these results, it can be concluded that the capability and reliability of MCNP-4C is constantly high in prediction of criticality accuracy for uranyl nitrate solution at STACY 280T slab core.
Currently, Indonesia through BATAN is operating three research reactors, namely the RSG-GAS reactor with the power of 30 MWt at Puspiptek south Tangerang (the first criticality in 1987), the TRIGA ...2000 reactor with the power of 2 MW in Bandung which the first criticality in 1965 with the power of 250 kW, was increased to 1 MW in 1971, and further upgraded to 2 MW in 2000. Beside that, there is Kartini reactor with a power of 100 kW located in Yogyakarta (first criticality in 1979). These reactors are quite old, and in accordance with Bapeten regulations, have carried out the first periodic safety review, to obtain a reactor license for the next 10 years of operation. In line with this, one of BATAN's current national research programs is to increase the production of radioisotopes and radiopharmaceuticals, where reactors play a very important role in the production of certain isotopes. In tracing the data obtained from operational reports related to irradiation requests from reactor users, namely PTRR, PSTNT, and PT INUKI for radioisotope production, which has been carried out in the last 5 years, May 2015 until 25 August 2020, show that the irradiation request at RSG-GAS is still not optimal. In term of the utilization of RSG-GAS, it can still be optimized, which in this case needs to be balanced with post-irradiation processing capabilities. Meanwhile, from the results of tracing and data collection, it can be shown that at this time the reactors are still operating. The utilization activities of the reactors complement each other according to their age and facilities.
In experimental power reactor (reaktor daya eksperimental, RDE), a full MOX fueled core can be accommodated in its operation without significant modifications on the reactor core design, but the MOX ...core tends to produce less favorable safety features and deliver transient behavior into unwanted fatal accidents. This paper aimed to investigate the MOX core characteristics of RDE through a series of calculations with MCNP6 code and ENDF/B-VII library. The calculation results show that MOX core with 235U enrichment of above 5 % can reach the criticality condition, maintain and run the reactor during the operation cycle. Utilizing MOX fuel in RDE with lower 235U enrichment will have a more negative impact on temperature coefficient of reactivity. The lower βeff makes the MOX core is more difficult to control, especially with low 235U enrichment. These results conclude that the selection of 235U enrichment in the MOX core must be carefully considered because it is one of the strategies to ensure the reactor safety criteria are fulfilled.
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