Articles
<![CDATA[Simulation of Void Detection System using Gamma-Ray Compton Scattering Technique ]]>
<![CDATA[Wirawan, Rahadi]]>;
<![CDATA[Waris, Abdul]]>;
<![CDATA[Kim, Hong Joo]]>;
<![CDATA[Djamal, Mitra]]>;
<![CDATA[Sari, Mona Berlian]]>
<![CDATA[ Journal of Engineering and Technological Sciences Vol 51, No 3 (2019) ]]>
Publisher : <![CDATA[ITB Journal Publisher, LPPM ITB ]]>
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DOI: 10.5614/j.eng.technol.sci.2019.51.3.5
<![CDATA[A simple void detection system for concrete was successfully developed using high-penetration gamma rays with Compton scattering. This research attempted to identify a void in the subsurface of a concrete volume that could not be accessed from any of the sides. Monte Carlo simulation using GEANT4 toolkit was performed to investigate the gamma-ray backscattering events. An NaI(Tl) detector was used with 60Co and 137Cs as gamma-ray sources. The void?s location was successfully detected during material target scanning. Density discrepancies conduce variance of the backscattering peak produced due to the presence of a void. Compared to 60Co as the gamma-ray source, 137Cs is a better choice for application in NDT systems using Compton scattering.]]>
<![CDATA[STUDI TENTANG FISIBILITAS DAUR ULANG AKTINIDA MINOR DALAM BWR ]]>
<![CDATA[Abdul Waris]]>;
<![CDATA[Budiono Budiono]]>
<![CDATA[ Jurnal Sains dan Teknologi Nuklir Indonesia (Indonesian Journal of Nuclear Science and Technology) Vol 9, No 1 (2008): Februari 2008 ]]>
Publisher : <![CDATA[BATAN ]]>
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DOI: 10.17146/jstni.2008.9.1.1829
<![CDATA[Studi awal tentang fisibilitas daur ulang aktinida minor (MA) tanpa melibatkan plutonium dalam reaktor air mendidih (boiling water reactor, BWR) telah dilakukan. Hasil yang diperoleh menunjukkan bahwa dengan menaikkan konsentrasi fisil MA dalam bahan bakar mixed oxide (MOX) dan/atau memperkecil fraksi hampa maka faktor multiplikasi efektif pada awal siklus akan naik juga, tetapi kenaikan ini belum cukup untuk membuat reaktor kritis. Disisi lain memperkecil void fraction akan menaikkan ayunan reaktivitas yang pada akhirnya akan menurunkan faktor keselamatan dari reaktor. ]]>
<![CDATA[Simulation of Void Detection System using Gamma-Ray Compton Scattering Technique ]]>
<![CDATA[Mona Berlian Sari]]>;
<![CDATA[Rahadi Wirawan]]>;
<![CDATA[Abdul Waris]]>;
<![CDATA[Hong Joo Kim]]>;
<![CDATA[Mitra Djamal]]>
<![CDATA[ Journal of Engineering and Technological Sciences Vol. 51 No. 3 (2019) ]]>
Publisher : <![CDATA[Institute for Research and Community Services, Institut Teknologi Bandung ]]>
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DOI: 10.5614/j.eng.technol.sci.2019.51.3.5
<![CDATA[A simple void detection system for concrete was successfully developed using high-penetration gamma rays with Compton scattering. This research attempted to identify a void in the subsurface of a concrete volume that could not be accessed from any of the sides. Monte Carlo simulation using GEANT4 toolkit was performed to investigate the gamma-ray backscattering events. An NaI(Tl) detector was used with 60Co and 137Cs as gamma-ray sources. The void's location was successfully detected during material target scanning. Density discrepancies conduce variance of the backscattering peak produced due to the presence of a void. Compared to 60Co as the gamma-ray source, 137Cs is a better choice for application in NDT systems using Compton scattering.]]>
<![CDATA[STUDI DESAIN REAKTOR AIR BERTEKANAN (PWR) BERUKURAN KECIL BERUMUR PANJANG BERBAHAN BAKAR THORIUM ]]>
<![CDATA[Moh Nurul Subkhi]]>;
<![CDATA[Zaki Suud]]>;
<![CDATA[Abdul Waris]]>;
<![CDATA[Sidik Permana]]>
<![CDATA[ JURNAL ISTEK Vol 9, No 1 (2015): ISTEK ]]>
Publisher : <![CDATA[JURNAL ISTEK ]]>
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<![CDATA[Penelitian desain small long-life PWR berukuran kecil, berumur panjang, berbahan bakar thorium sudah berhasil dilakukan. Riset ini merupakan kelanjutan riset sebelumnya dengan melakukan survei parameter dan optimasi untuk disain teras pada daya sebesar 350 MWt. Teras didisain agar dapat beroperasi secara kritis 10 tahun tanpa pengisian bahan bakar. Reaktor termal kecil dapat beroperasikan dengan waktu yang lama tanpa refueling denga cara memilih bahan bakar dengan internal conversion ratio besar dan ekses reaktivitas rendah selama waktu operasi. Oleh karenanya ada tiga strategi yang akan dilakukan seperti memanfaatkan thorium cycle, memakai konsep tight lattice dan menambahkkan burnable poisson Protactinium-231 dalam bahan bakar. Pada penelitian ini akan dilakukan survei parameter untuk fraksi bahan bakar dari 35% sampai dengan 65%, konsentrasi U-233 dan Pa-231 yang variatif dan optimasi neutronik juga akan dilakukan untuk mereduksi ukuran geometri teras aktifekses reaktivitas seoptimal mungkin dengan distribusi daya yang merata. Perhitungan cell dan difusi multigrup pada penelitian ini dilakukan dengan memanfaatkan kode SRAC-COREBN yang dikembangkan oleh JAERI dengan memanfaatkan data nuklida JENDL-3.2/3.3.]]>
<![CDATA[Plutonium and Minor Actinides Recycling in Standard BWR using Equilibrium Burnup Model ]]>
<![CDATA[Abdul Waris]]>;
<![CDATA[Rizal Kurniadi]]>;
<![CDATA[Zaki Su'ud]]>
<![CDATA[ Journal of Mathematical and Fundamental Sciences Vol. 40 No. 1 (2008) ]]>
Publisher : <![CDATA[Institute for Research and Community Services (LPPM) ITB ]]>
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DOI: 10.5614/itbj.sci.2008.40.1.2
<![CDATA[Plutonium (Pu) and minor actinides (MA) recycling in standard BWR with equilibrium burnup model has been studied. We considered the equilibrium burnup model as a simple time independent burnup method, which can manage all possible produced nuclides in any nuclear system. The equilibrium burnup code was bundled with a SRAC cell-calculation code to become a coupled cell-burnup calculation code system. The results show that the uranium enrichment for the criticality of the reactor, the amount of loaded fuel and the required natural uranium supply per year decrease for the Pu recycling and even much lower for the Pu & MA recycling case compared to those of the standard once-through BWR case. The neutron spectra become harder with the increasing number of recycled heavy nuclides in the reactor core. The total fissile rises from 4.77% of the total nuclides number density in the reactor core for the standard once-through BWR case to 6.64% and 6.72% for the Plutonium recycling case and the Pu & MA recycling case, respectively. The two later data may become the main basis why the required uranium enrichment declines and consequently diminishes the annual loaded fuel and the required natural uranium supply. All these facts demonstrate the advantage of plutonium and minor actinides recycling in BWR.]]>
<![CDATA[Feasibility Study on Long-Lived Fission Products Transmutation in Equilibrium Fuel Cycles of PWR ]]>
<![CDATA[Abdul Waris]]>;
<![CDATA[H. Sekimoto]]>
<![CDATA[ Indonesian Journal of Physics Vol 16 No 4 (2005): Vol. 16 No. 4, October 2005 ]]>
Publisher : <![CDATA[Institut Teknologi Bandung ]]>
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<![CDATA[A study on long-lived fission products (LLFP) transmutation in equilibrium fuel cycles of pressurized water reactors (PWR) has been performed. The seven important LLFP, i.e., 79Se, 93Zr, 99Tc, 107Pd, 126Sn, 129I, and 135 Cs were considered in this study. In order to get a comprehensive perspective, single isotope/element transmutations and mixed isotopes/elements transmutations for both isotopic separation and elemental separation were employed. The calculation results showed that the transmutation of the all seven LLFP together with isotopic separation are feasible by increasing the required enrichment of loaded uranium fuel up to 7.1wt% for standard PWR. Isotopic separation is not necessary for mixed technetium and iodine transmutation. The transmutation of the all seven LLFP together with elemental separation may not possible since zirconium transmutation in PWR is a formidable task.]]>
<![CDATA[Design of The 200 MW(t) Pb-Bi Cooled Nuclear Fast Reactor For Steam Membrane Reforming Hydrogen Production ]]>
<![CDATA[Epung Saepul Bahrum]]>;
<![CDATA[Zaki Su’ud]]>;
<![CDATA[Abdul Waris]]>;
<![CDATA[Bambang Ari Wahjoedi]]>
<![CDATA[ Indonesian Journal of Physics Vol 17 No 4 (2006): Vol. 17 No. 4, October 2006 ]]>
Publisher : <![CDATA[Institut Teknologi Bandung ]]>
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<![CDATA[Design study of the 200 MW(t) Pb-Bi cooled nuclear fast reactor for steam membrane reforming hydrogen production has been proposed. The design base on neutronic, steady state thermalhydraulic and safety examination of the 200 MWt Pb-Bi cooled fast nuclear reactor. For balance cylindrical core demonstrated that at flowrate 7000 kg s-1 and inlet coolant temperature of 350 0C yield average and maximum coolant outlet temperature of 552 and 621 0C. The safety analysis confirm at that situation the coolant and fuel cladding maximum temperature lower than coolant boiling and cladding melting temperature. The 552 0C average coolant outlet temperature corresponds to maximum reaction temperature of the steam membrane reforming which is close to 550 0C. We proposed design of the 200 MW(t) Pb-Bi cooled balance cylindrical core nuclear fast reactor for steam membrane reforming hydrogen production. The configuration of the design is Pb-Bi cooled nuclear fast reactor and steam membrane hydrogen production unit coupled with intermediate heat exchanger. One of the intermediate heat exchanger functions is to protect nuclear reactor when an explosion accident occur in the hydrogen production unit.]]>
<![CDATA[Feasibility Study of Integrating Absorption Heat Pump into Methanol Steam Reforming Process for Hydrogen Production ]]>
<![CDATA[Willy Yanto Wijaya]]>;
<![CDATA[Ken Okazaki]]>;
<![CDATA[Kazuyoshi Fushinobu]]>;
<![CDATA[Abdul Waris]]>
<![CDATA[ Indonesian Journal of Physics Vol 18 No 3 (2007): Vol. 18 No. 3 July 2007 ]]>
Publisher : <![CDATA[Institut Teknologi Bandung ]]>
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<![CDATA[Theoretically, Methanol Steam Reforming (MSR) process to produce hydrogen only requires endothermic heat with temperature lower than 100°C. Meanwhile, waste heat in the temperature level of 100-150°C is disposed in huge amount by various industrial sectors every year. If this abundant waste heat can be recovered and stored into hydrogen energy through the MSR, a potential gain and high-efficient energy system could be achieved. However, empirically, temperature level ≥ 200°C is required for the MSR process to have high conversion from the methanol to hydrogen. Therefore, Absorption Heat Pump (AHP) system is then utilized to enhance the temperature level of the waste heat. Nevertheless, AHP system certainly requires additional input energy. This research was to investigate the feasibility of integrating the AHP system into MSR process to produce hydrogen. Further consideration on hydrogen to electricity conversion using Fuel Cell (FC) would also be provided. The feasibility study was conducted through the efficiency calculations, either for energy or exergy terms, for several theoretical as well as actual-approximation cases.]]>
<![CDATA[The New Fugacity Calculation in Finite Nuclear Matter ]]>
<![CDATA[Rizal Kurniadi]]>;
<![CDATA[Abdul Waris]]>
<![CDATA[ Indonesian Journal of Physics Vol 18 No 3 (2007): Vol. 18 No. 3 July 2007 ]]>
Publisher : <![CDATA[Institut Teknologi Bandung ]]>
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<![CDATA[The caloric curve of hot nuclei in equilibrium states have been calculated with a new prescription namely a new fugacity calculation. In this technique, the fugacity is directly proportional to the nucleon density; hence, the fugacity is obtained through thermal wavelength. In contrast with the constant fugacity, the choosing of thermal wavelength approximation gives a simpler way to calculate the density profile and the entropy of finite nuclear matter. Variation of thermal wavelength value does not affect to the density and entropy. The phase transition temperature is dependent concerning both of the thermal wavelength value and the potential deep.]]>
<![CDATA[Thorium Fuel Cycle in BWR with Free 233U ]]>
<![CDATA[Abdul Waris]]>;
<![CDATA[W. Nawangsary]]>;
<![CDATA[Syeilendra Pramuditya]]>;
<![CDATA[Rizal Kurniadi]]>
<![CDATA[ Indonesian Journal of Physics Vol 18 No 4 (2007): Vol. 18 No. 4, October 2007 ]]>
Publisher : <![CDATA[Institut Teknologi Bandung ]]>
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<![CDATA[Study on thorium fuel cycle with free 233U in boiling water reactor (BWR) has been carried out. In this study we have utilized plutonium and minor actinides (MA) as fissile nuclides instead of 233U as one of the main scenario to obtain the 233U free BWR core with thorium. Beside that, the void-fraction of the reactor is modified from 20% up to 70%. The results show that the standard BWR core can maintain its criticality when the loaded fuel is thorium with 11.16% and 1.24 % of plutonium and MA, respectively. The use of 11.16% of Pu and 1.24 % of MA is more than enough to substitute the 1% of 233U to obtain thorium fuel cycle in the standard BWR with free 233U. Moreover, the lesser amount of plutonium and minor actinides in the fuel will results in the great degradation on the safety of reactor]]>
