Papers by Alexander Agung
Indonesian Journal of Physics and Nuclear Applications
Simulation using Monte Carlo code has been conducted to determine the distribution of absorbed do... more Simulation using Monte Carlo code has been conducted to determine the distribution of absorbed dose to the breast brachytherapy with 131Cs and 103Pd radionuclide sources. Simulations performed on stage I breast cancer with cancer diameter is 2 cm. Sources of radionuclides simulated in the form of seed is modeled with CS-1 which is made by IsoRay 131Cs and seed ADVANTAGETM103Pd which is made by IsoAID, LLC. Seed was planted in breast cancer cells. Calculation of absorbed dose distribution was performed by varying the distance from the seed. Variations of the distance started from a radius of 0.3 cm to 2 cm with a range of 0.1 cm respectively. In this simulation will also be reviewed the value of absorbed dose for healthy cell like breast, sternum, and lung. The relation between the absorbed dose and the distance from the seed can be described in the form of power law. The results of the calculation show that the maximum absorbed dose is in the target site of the cancer cells (5.791 ±...
IOP Conference Series: Earth and Environmental Science
Fuel loading pattern optimization is a complex problem because there are so many possibilities fo... more Fuel loading pattern optimization is a complex problem because there are so many possibilities for combinatorial solutions, and it will take time to try it one by one. Therefore, the Polar Bear Optimization Algorithm was applied to find an optimum PWR loading pattern based on BEAVRS. The desired new fuel loading pattern is the one that has the minimum Power Peaking Factor (PPF) value without compromising the operating time. Operating time is proportional to the multiplication factor (k eff ). These parameters are usually contradictive with each other and will make it hard to find the optimum solution. The reactor was modelled with the Standard Reactor Analysis Code (SRAC) 2006. Fuel pins and fuel assemblies are modelled with the PIJ module for cell calculations. One-fourth symmetry was used with the CITATION X-Y module for core calculations. The optimization was done with 200 populations and 50 iterations. The PPF value for the selected solution should never exceed 2.0 in every burn...
Masyarakat Ekonomi ASEAN memberikan peluang pasar bebas di ranah ASEAN. Biaya produksi listrik ya... more Masyarakat Ekonomi ASEAN memberikan peluang pasar bebas di ranah ASEAN. Biaya produksi listrik yang rendah akan meningkatkan daya saing produk. Salah satu solusinya ialah membangun Pembangkit Listrik Tenaga Nuklir (PLTN). Faktor keselamatan, keamanan, dan efisiensi menjadi pertimbangan untuk menentukan jenis reaktor nuklir yang akan dibangun. Salah satu jenis reaktor yang menarik untuk dikembangkan adalah High Temperature Gas-cooled Reactor (HTGR) pebble bed. Bahan bakar HTGR yang digunakan pada penelitian ini adalah uranium oksida (UO2). Penelitian ini bertujuan untuk mendapatkan desain teras reaktor yang optimal pada daya 150 MWt. Untuk mencapai tujuan tersebut, dilakukan analisis neutronik pada kondisi teras ekulibrium sehingga didapatkan parameter-parameter teras optimal, yaitu nilai burn up, distirbusi daya, dan fuel residence time. Penelitian ini memvariasikan geometri teras dan pengayaan bahan bakar. Radius teras berkisar dari 1 hingga 1,5 meter dengan kenaikan 0,1 meter, sed...
Banyaknya kombinasi peletakan susunan perangkat bahan bakar di dalam teras diawal operasi reaktor... more Banyaknya kombinasi peletakan susunan perangkat bahan bakar di dalam teras diawal operasi reaktor, maka perlu dilakukan optimasi agar dapat diperoleh konfigurasi teras yang optimum dengan nilai keff akhir siklus yang maksimum dan nilai faktor daya puncak (PPF) yang minimum. Terdapat dua metode Genetic Algorithm yang digunakan dalam optimasi ini yaitu objektif tunggal dan multiobjektif. Optimasi dilakukan pada model 1⁄4 simetri teras (52 posisi perangkat bahan bakar) dengan 3 tipe perangkat bahan bakar yaitu perangkat dengan pengkayaan U-235 sebesar 1,5% sebanyak 13 buah, 2,5% sebanyak 15 buah dan 3% sebanyak 24 buah tanpa batang racun dapat bakar. Perhitungan neutronik tingkat perangkat bahan bakar menggunakan kode PIJburn, sedangkan tingkat teras menggunakan kode COREBN. Dari optimasi objektif tunggal didapatkan konfigurasi optimum dengan perpanjangan cycle length sebesar 8,9% (60 hari) dan penurunan PPF sebesar 23,31% terhadap konfigurasi model standar. Sedangkan dari optimasi mul...
Neutronic calculations have been performed to a fluidized bed nuclear reactor that uses advanced ... more Neutronic calculations have been performed to a fluidized bed nuclear reactor that uses advanced coated particles to improve its endurance against irradiation and high temperature. The calculation is intended to determine whether reactor characteristics have been significantly compromised. The characteristic of the reactor is assessed by investigating the change in criticality from packed bed condition to a full expansion of the particle bed. The particle used in this calculation was based on the standard TRISO fuel particle as being used in the HTR-10 reactor, and the use of advanced fuel particle was performed by replacing the SiC layer by a ZrC layer. A packed bed of 50 cm high was used in this research with an additional 20 ppm of boron in the side reflector. At packed condition, replacing SiC by ZrC in TRISO particles significantly decreases the criticality with a range of -336 ± 138 pcm to -2809 ± 99 pcm. Calculation on expanded bed shows similar behaviors, in which case the d...
Reaktor Kartini telah beroperasi hampir 40 tahun. Perpanjangan izin pada tahun 2019 memerlukan an... more Reaktor Kartini telah beroperasi hampir 40 tahun. Perpanjangan izin pada tahun 2019 memerlukan analisa yang lebih mendalam terkait dengan persyaratan keselamatan dari Bapeten. Peraturan terbaru dari badan pengawas tahun 2014 mensyaratkan perhitungan komputasional untuk parameter keselamatan teras. Dalam penelitian ini Software TRIGA MCNP telah digunakan reaktor Kartini untuk data pembanding sekaligus untuk validasi pemodelan komputasional. Hasil perhitungan komputasional telah dibandingkan dengan hasil pengukuran secara eksperimental dalam kurun waktu 8 tahun terakhir. Secara umum reaktivitas lebih dan reaktivitas margin padam teras reaktor Kartini sesuai dan memenuhi Batasan Kondisi Operasi (BKO) yang telah ditetapkan dalam Laporan Analisis Keselamatan (LAK). Hasil analisastatistik ANOVA memberikan kesimpulan bahwa model komputasional reaktivitas lebih teras belum memberikan hasil yang sama dengan data eksperimentalnya. Tren grafik menunjukkan kemiripan dengan perbedaan varian yang...
The Kartini Triga Mark II Research Reactor (Kartini TM2RR), one of the three research reactors in... more The Kartini Triga Mark II Research Reactor (Kartini TM2RR), one of the three research reactors in Indonesia, is a pool-type reactor with a thermal power of 100 kW that has been operating since 1979 and has legal operation licensing until 2019 from the regulatory body. As of 2018, the reactor is composed of 71 cylindrical fuel elements of height 66 cm in a hexagonal array, with most of the fuel elements remaining untouched from their first loading in the reactor in 1994. The reactor also comprises three rods for reactivity control and nine peripheral graphite block elements. The reactor is operated on a per-user-demand basis with irregular cycle lengths (at the maximum six hours operation per day) with an average operation time of 200 hours a year. A core safety analysis of the Kartini TM2RR is currently being prepared to complete requirements from the regulatory body. In this safety analysis, core excess reactivity and reactivity shutdown margin are the main parameters to be calcula...
Salah satu komponen penting dalam pembangkitan daya pada reaktor nuklir adalah steam generator. H... more Salah satu komponen penting dalam pembangkitan daya pada reaktor nuklir adalah steam generator. High Temperature Gas-Cooled Nuclear Reactor (HTGR) merupakan salah satu reaktor nuklir dengan suhu keluaran yang tinggi serta menggunakan steam generator berjenis once through helical coil. Salah satu potensi kecelakaan dasar desain pada HTGR adalah Depressurized Loss of Forced Cooling (DLOFC) di mana tekanan pendingin reaktor turun secara signifikan akibat patahan pada saluran pendingin. Transien tekanan dan aliran pendingin ini akan mempengaruhi perilaku dinamik steam generator. Program RELAP5-3D digunakan untuk mensimulasikan respon kejadian transien pada steam generator reaktor HTR berdaya 150 MWt berdasar pada hukum konservasi massa, momentum dan energi yang dirumuskan sebagai two-fluid model beserta relasi-relasi penting hidrodinamika yang lain. Hasil simulasi menunjukan bahwa penurunan laju aliran massa pada sistem primer akan mengakibatkan daya, suhu pendingin primer, suhu pending...
FLUBER (Fluidized Bed Thermal Fission Nuclear Reactor) is a conceptual design of a modular reacto... more FLUBER (Fluidized Bed Thermal Fission Nuclear Reactor) is a conceptual design of a modular reactor utilizing the concept of fuel pellet suspension. It consists of TRISO coated fuel particles contained in a graphite-walled cylinder. Helium is used as a coolant and as fluidizing medium. It has been observed experimentally that particles in a gas-solid fluidized bed move chaotically and the flow structure is characterized by the occurrence of void regions (or bubbles). The bubble formation and accompanying fuel particle movement present an inhomogeneous state of fuel particle distribution in the core, affecting the reactivity of the core. To investigate the influence of bubble formation on reactivity of the new design, some static calculations were performed using KENO-V.a code. The reactivity of the inhomogeneous core is compared with that of the corresponding homogeneous core. Further, a theoretical model describing the coupling of neutronics, thermohydraulics and fluidization in a f...
Research within nuclear safety and radiation protection is necessary in order to maintain the hig... more Research within nuclear safety and radiation protection is necessary in order to maintain the high level of competence required by an expert authority. In the field of reactor safety research, SSM’s goals are to support regulation and contribute to national competence in the area of nuclear safety. A technical support organization on deterministic safety analysis (TSO-DSA) has been set up to help SSM in fulfilling these goals. The TSO-DSA function was then established by SSM at two nuclear universities, i.e. Royal Institute of Technology (KTH) in Stockholm and Chalmers University of Technology in Gothenburg. Activities related to this function have been performed, emphasizing the use of best-estimate coupled codes (i.e. PARCS/RELAP5 and PARCS/TRACE) for the analyse . The activities performed by Chalmers are reported in this paper as examples. The on-going activities give a good example on how the safety authority co-operates with universities. The use of coupled codes gives satisfac...
National Atomic Energy Agency of Indonesia (BATAN) has chosen the High Temperature Gas-cooled Rea... more National Atomic Energy Agency of Indonesia (BATAN) has chosen the High Temperature Gas-cooled Reactor (HTGR) for energy fulfilment solution in Indonesia. HTGR is a high-level safety Gen. IV power reactor, no melting core when an accident occurs, which very suitable for Indonesia. Indonesian HTGR development is initiated by the development of its experimental type, named Reaktor Daya Eksperimental (RDE). RDE refers to the Chinese HTR-10 design that had reached full power operation in 2003. Various researches have been conducted to prepare the RDE design to meet the HTGR safety system. This research is aimed to understand the characteristics of primary fluid flow and heat transfer in the core of HTR-10 and in that respect it is possible to obtain important parameters that can be used in RDE design. HTR-10 core modelling, research methods, and the results are described herein.
International Journal on Advanced Science, Engineering and Information Technology
KLT-40S nuclear reactor is a small modular floating nuclear power plant made by Russia as a conve... more KLT-40S nuclear reactor is a small modular floating nuclear power plant made by Russia as a conventional light water reactor (LWR) problem solution nowadays, such as high overnight cost, long commissioning period, and lack of flexibility in supplying a small load of electricity and supplying electricity to isolated areas. With those characteristics, the KLT-40S is suitable to be applied to isolated areas with a small electrical load like archipelagic states such as Indonesia. Based on that reason, Indonesia needs to assess the KLT-40S floating nuclear power plant feasibility study through explorative research. One of those is assessing the reactor core neutronic parameter. In this research, the reactor core modelling is done by using the KENO VIA and T-6DEPL module in SCALE 6.1 code package. Several variations of reactor operating parameters such as fuel composition and configuration, fuel temperature, moderator temperature, and moderator void fraction had been done in this research. The aim was to get several neutronic parameters to confirm the core feasibility from operational and inherent safety perspectives. Those neutronic parameters are fuel cycle length and reactivity feedback coefficient of fuel temperature, moderator temperature, and moderator void fraction. Based on this research result, the fuel configuration that produces 28 months of cycle length is the fuel base of dispersed UO 2 in the silumin matrix with 18,6 % 235 U enrichment. Both of the two fuel bases used in this research have inherent safety characteristics, which are shown by the negative value of the reactivity feedback coefficient of fuel temperature, moderator temperature, and moderator void fraction. Dispersed UO 2 in the silumin matrix fuel base has better inherent safety characteristics than the UO 2 ceramic metal fuel base.
International Journal on Advanced Science, Engineering and Information Technology
One of Floating Nuclear Power Plant (FNPP) designs in the world is currently being built by Russi... more One of Floating Nuclear Power Plant (FNPP) designs in the world is currently being built by Russian Federation, named “Academic Lomonosov,” which uses two PWR types, KLT-40S as its power unit. However, too little information regarding its detailed technical specification is available, including its thermal-hydraulics parameters. The objective of this research is to create a thermal-hydraulic model of KLT-40S reactor core use RELAP5-3D and to predict fuel and cladding temperature value at the steady-state condition, and transient condition with a variety of primary coolant mass flow rate and pressure to simulate abnormal event within the reactor. The reactor thermal-hydraulic model is created by dividing 121 coolant channels in the actual nuclear fuel assemblies into two channels: one channel to simulate coolant flow in 120 fuel assemblies with average heat generation, and the other channel to simulate coolant flow in one fuel assembly with highest heat generation in the core. The fuel structure had solid cylinder geometry and made from ceramic-metal UO2 dispersed in the inert silumin matrix. The fuel cladding is made from zirconium alloy. These fuel heat structures generate heat from fission reaction and are modelled as a heat source according to the reactor power technical data, i.e., 150 MWt. The reactor axial power distribution is approximated by cosine distribution. Operation parameter variation that represents the real reactor normal operation condition in this research is a variation that has flow loss coefficient value 8,000, radial power peaking factor 1.1, and axial power peaking factor 1.1 with axial power peaking located in the middle of the fuel rod. The fuel and cladding temperature value at the steady-state condition and several transient conditions are predicted in this research, and there is no temperature value that goes beyond the safety limit.
International Journal on Advanced Science, Engineering and Information Technology, Jun 8, 2018
A study about Resource Renewable Boiling Water Reactor (RBWR) core, a reduced moderation boiling ... more A study about Resource Renewable Boiling Water Reactor (RBWR) core, a reduced moderation boiling water reactor that features the breeding ratio larger than 1 was conducted. This study focuses on the neutronic performances of the core and aims to investigate the core sustainability when using thorium as the main fertile fuel. A fuel-self-sustaining core with high burnup set as the design target. 233 U+Th were used as the initial fuel, and the impact of initial fissile (233 U) content in the core fissile zone on the core neutronic performances was evaluated. Parameters related to the neutronic performances such as the core burnup, fissile breeding, and fissile inventory ratio (FIR) are considered in this study. From these results, it was confirmed that it is feasible to create a selfsustaining fuel cycle system using thorium fueled RBWR. However, there was a trade-off between the core burnup and fissile breeding that can be a significant challenge in the development of this system. Evaluating the other design variables may be considered to address this challenge. The further study to analyze the safety performances of the core is required to arrive at a safe and reliable reactor system.
International Journal of Nuclear Energy Science and Technology
Safety analysis of a PWR fuelled with ATF (Accident-Tolerant Fuel) has been performed at LB-LOCA ... more Safety analysis of a PWR fuelled with ATF (Accident-Tolerant Fuel) has been performed at LB-LOCA condition. The ATF being used is uranium silicide (U3Si2) and FCMF (Fully Ceramic Microencapsulated Fuel) with silicon carbide (SiC) and FeCrAl alloy as a cladding material. The objective of this research is to obtain dynamic characteristics of ATF-fuelled PWR at LB-LOCA condition. RELAP5-3D system code was used to model the reactor and simulate the transient. A safe shutdown of the reactor was assumed after a depressurisation following a double-ended guillotine breach in the main pipe. The results of simulations show that during LB-LOCA with partially functioning ECCS, the transient PCTs were far below the maximum allowable limit. The use of ATF could decrease the maximum transient PCT. It is shown that U3Si2 fuel with FeCrAl cladding has the minimum PCT transient and the shortest quench time to steady state condition after transient initiation.
INTERNATIONAL ENERGY CONFERENCE ASTECHNOVA 2019
TerraPower Traveling Wave Reactor-Prototype (TWR-P) provides the potential to properly address de... more TerraPower Traveling Wave Reactor-Prototype (TWR-P) provides the potential to properly address depleted uranium waste problems through an improved uranium utilization system. Despite all the economic and environmental benefits, the concept is shrouded with criticism against its use of sodium as the coolant due to the experience of sodium leaks and its unwary design to enrichment. The sodium-cooled fast reactor has a checkered history, with accidents such as Superphenix (France) and Monju (Japan). This is because sodium characteristic that burns on contact with air and reacts violently with water when leaks occurred in the reactor. Such an accident is often followed by expensive processes of cleanup, testing, and inspection before the reactor able to be restarted. Thus, the paper pursues an alternative to sodium, by looking at the temperature coefficient, initial multiplication factor, and conversion ratio. In addition to that, the paper further discusses possible loading pattern modifications to improve t...
INTERNATIONAL ENERGY CONFERENCE ASTECHNOVA 2019
NuScale is an integral UO2 fueled PWR that operates with natural circulation. In this study therm... more NuScale is an integral UO2 fueled PWR that operates with natural circulation. In this study thermal-hydraulic analysis was carried out on the NuScale primary system to observe the natural convection phenomenon when the fuel was changed to mixed oxide (MOX). The use of MOX could increase the neutron advantage, but it was offset by enlargement of the core diameter. This modification may cause the thermal properties of the fuel and the flow distribution in the core changes. It is thus necessary to analyze the effect of such changes to ensure that NuScale's natural convection capability is maintained and the reactor remains safe. This research was carried out using the RELAP5-3D thermal-hydraulics code. Four thermal-hydraulics models were analyzed, based on the properties of high burn-up fuel. Each type of fuel was simulated under end-of-cycle (EOC) and beginning-of-cycle (BOC) conditions. BOC simulations were used to test reactor safety when operating with a large power peaking factor (PPF), while the fuel properties were maintained in high burn-up conditions. The results showed that the ability of natural circulation in each model remained able to be maintained based on differences in cooling density. There were no significant differences in the coolant temperature, cooling flow rate, and void fraction of each model. The enlargement of the core diameter, however, causes an increase in the fuel channel void fraction due to the reduction of cooling flow within. A significant impact occurred at the pellet temperature, where the highest pellet temperature occurred at MOX because MOX conductivity was lower than UO2, but the pellet peak temperature was below the melting temperature of MOX. There are no operating parameters that exceed the safety limit, so the reactor can still maintain its natural convection capability, and MOX is suitable for use at the NuScale reactor.
INTERNATIONAL ENERGY CONFERENCE ASTECHNOVA 2019
KLT-40S has an advantage as a floating nuclear power plant with high mobility and portability bec... more KLT-40S has an advantage as a floating nuclear power plant with high mobility and portability because it can be towed throughout the sea to the destination. The changes in the sea conditions when the nuclear reactor operates should be accounted for the thermal-hydraulic analysis. The sea state condition and the reactor power were varied to know how the characteristics of the thermal-hydraulic parameters at various sea state conditions, especially the void fraction. RELAP5-3D was used to analyze the thermal-hydraulic characteristics of the floating nuclear power plant in two different types of reactor motion with nine sea state conditions. The KLT-40S reactor core model includes peak channel, average channel, and bypass channel is developed in RELAP5-3D code. The result of the analysis shows that the effect of reactor motion is higher in the pitching – heaving motion than the rolling – heaving motion. In extreme sea state condition of the pitching – heaving motion, the void formation is started from 70% of reactor power. The highest void fraction observed in this condition with 110% of reactor power is 5.858 %. Otherwise, in extreme sea state condition of the rolling – heaving motion, the void formation is started from 100% of reactor power. The highest void fraction observed in this condition with 110% of reactor power is 0.07%.
E3S Web of Conferences
Nuclear fuel management was done by optimizing fuel loading pattern in a reactor core. Practicall... more Nuclear fuel management was done by optimizing fuel loading pattern in a reactor core. Practically, performing fuel loading pattern optimization was difficult because of its combinatorial problem complexity which needed to be solved. Therefore, Quantum-inspired Evolutionary Algorithm (QEA) which could solve the combinatorial problem faster than conventional method was used. The main purpose of this research was to obtain an optimum fuel loading pattern of KSNP-1000 reactor core without altering fuel assembly inventories. KSNP-1000 core was modeled in SRAC code package using PIJ module for fuel pins and fuel assemblies’ lattices and CITATION module for fuel assemblies’ pattern in a quarter core symmetry. Optimization problem adaptation using QEA was made by presenting 52 fuel assemblies in Q-bit individuals with the length of 8 Q-bits. Q-bits were converted to corresponding bit values and then given weight which would be used as consideration to optimize the pattern. The optimization...
Progress in Nuclear Energy
Abstract Space exploration is very important for the future of the earth and human beings as it m... more Abstract Space exploration is very important for the future of the earth and human beings as it may eliminate earth overpopulation and overcome diminishing of earth resources. One of the obstacles of the space exploration mission is the energy source for the spacecraft. One alternative is using a nuclear reactor as an energy source in spacecraft. A conceptual design of Indonesian Space Reactor (ISR) has been carried out to explore such a possibility. ISR is a liquid metal Na-78 K cooled space reactor with a fast neutron spectrum. It is designed to provide at least 500 kWth power for operating time more than 10 years at full power. The reactor uses 55% high-enriched uranium nitrate as fuel. The ISR hexagonal core is comprised of 61 fuel pins and is designed in the form of a hollow cylinder with an individual cooling channel in each fuel pin. The reactor is also equipped with spectral shift absorbers (SSA) made of Re and Mo-30Re alloy to control the reactivity. Neutronic calculations have been performed to obtain optimum design parameters without compromising safety requirements. These design parameters include variation in uranium enrichment, reactor dimension, reflector thickness and control drum (absorber) design and dimension. The accepted reactor design has an excess reactivity of 4023 ± 9 pcm and shutdown margin of 4852 ± 9 pcm and the reactor is estimated to have a lifetime of 28 years. The temperature and void reactivity coefficients are all negative, implying inherent safety. Several accident scenarios were also considered in this work, both during launch failure and normal operation. It is found that to keep the reactor subcritical for a submerged reactor following a launch failure, the reflector segment should be discarded. Meanwhile, some portions of fuel pins should be removed from the core during operational accidents.
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Papers by Alexander Agung