Papers by Tagor Sembiring
An optimum fuel composition is a very important parameter in the operation of a pebble bed high t... more An optimum fuel composition is a very important parameter in the operation of a pebble bed high temperature gas-cooled reactor (HTGR). In the present scoping study, the optimum ranges of heavy metal (HM) loading per pebble and the uranium enrichment are investigated. The HM loading range covers 4 to 10 g per pebble, while the uranium enrichment covers 5 to 20 w/o. Two fuel loading schemes typical to pebble-bed HTGRs are also investigated, i.e. the OTTO and multi-pass schemes. All calculations are carried out using BATAN-MPASS, a general in-core fuel management code dedicated for pebble-bed type HTGRs. The reference reactor design case is adopted from the German 200 MWth HTR-Module but with core height of half of the original design. Other design parameters follow the original HTR-Module design. The results of the scoping study show that, for both once-through-then-out (OTTO) and multi-pass fueling schemes, the optimal HM loading per pebble is around 7 g HM/ball. Increasing the urani...
Reactivity insertion accident analysis during uranium foil target irradiation in the RSG-GAS reactor core
Nuclear Technology & Radiation Protection, 2020
Analysis of the steady-state and reactivity insertion accident is very important for the safety o... more Analysis of the steady-state and reactivity insertion accident is very important for the safety of reactor operations. In this study, steady-state and reactivity insertion accident analysis when the low enriched uranium foil target is irradiated in the reactor core has been carried out. The analysis is carried out by the best estimate method by using a coupled neutronic, kinetic, and thermal-hydraulic code, MTR-DYN. The MTR-DYN code is based on the 3-D multigroup neutron diffusion method. The cell calculations for the target are carried out by the WIMSD/5 and MTR-DYN code. After reactivity insertion, the coolant, fuel, and clad temperature are observed. The calculation results for the initial power of 1 W showed that the maximum temperature of the coolant, clad, and fuel are 49.76?C, 65.01?C, and 65.26?C, respectively. Meanwhile, when the reactivity insertion at the initial power of 1 MW, the maximum temperature of the coolant, clad, and fuel are 72.23?C, 140.79?C, and 141.97?C, res...
Analysis of the excess reactivity and control rod worth of RSG-GAS equilibrium silicide core using Continuous-Energy Monte Carlo Serpent2 code
Annals of Nuclear Energy
Evaluation on fuel cycle and loading scheme of the Indonesian experimental power reactor (RDE) design
Nuclear Engineering and Design
THE 4TH BIOMEDICAL ENGINEERING’S RECENT PROGRESS IN BIOMATERIALS, DRUGS DEVELOPMENT, HEALTH, AND MEDICAL DEVICES: Proceedings of the International Symposium of Biomedical Engineering (ISBE) 2019
Silicon carbide (SiC) fuel cladding has several primary advantages over zircaloy claddings. SiC c... more Silicon carbide (SiC) fuel cladding has several primary advantages over zircaloy claddings. SiC cladding absorbs very low thermal neutrons, little corrosion, and no hydrogen pickup during normal operation, making it an attractive choice for use as cladding on PWR fuel. In this paper, AP1000 neutronic parameters are evaluated using SiC and ZIRLO (Westinghouse's zircaloy) cladding. The steady-state calculation for the fuel element was simulated by the PIJ module of SRAC2006 code as well as the core calculation using the NODAL3 code. The calculated kinf values are always higher for SiC cladding including all fuel temperature reactivity coefficient is negative same as ZIRLO cladding. The core calculation shows that the excess reactivity is higher 337 pcm using SiC cladding than ZIRLO cladding. It shows that the SiC cladding can be used in the AP1000 with some correction because of the quite higher excess reactivity.
THE 4TH BIOMEDICAL ENGINEERING’S RECENT PROGRESS IN BIOMATERIALS, DRUGS DEVELOPMENT, HEALTH, AND MEDICAL DEVICES: Proceedings of the International Symposium of Biomedical Engineering (ISBE) 2019
The in-core fuel management of a nuclear power plant is a problem of optimization of core paramet... more The in-core fuel management of a nuclear power plant is a problem of optimization of core parameters such as operation cycle, average fuel burnup and shut down margin for determining a fuel loading pattern to meet the safety and economic aspects. The study is aimed to obtain an optimal fuel loading pattern. Two models of fuel burn up calculations were taken namely equilibrium and transition cores burn up models. The calculations will be carried out by means of computer codes SRAC2006 for cell calculation and PWR-FUEL for the fuel management. The results of keff values at BOC and EOC for each transition core are approximately 1.05 as the input data and the core cycle length is found to be 330 days. The keff values at both BOC and EOC are very near to critical at equilibrium core and the core cycle length is found 360 days. The results of the calculation of neutron flux distribution and power density using the NODAL and FDM methods of the PWR-FUEL the code has the same results. From the results of the neutronic parameter, it is shown that the optimal loading pattern of PWR core can be determined by the PWR-FUEL code either with equilibrium core search or with transition core burnup models.
Journal of Physics: Conference Series
The reactivity coefficient is a very important parameter for inherent safety and stability of nuc... more The reactivity coefficient is a very important parameter for inherent safety and stability of nuclear reactors operation. To provide the safety analysis of the reactor, the calculation of changes in reactivity caused by temperature is necessary because it is related to the reactor operation. In this paper, the temperature reactivity coefficients of fuel and moderator of the AP1000 core are calculated, as well as the moderator density and boron concentration. All of these coefficients are calculated at the hot full power condition (HFP). All neutron diffusion constant as a function of temperature, water density and boron concentration were generated by the SRAC2006 code. The core calculations for determination of the reactivity coefficient parameter are done by using NODAL3 code. The calculation results show that the fuel temperature, moderator temperature and boron reactivity coefficients are in the range between-2.613 pcm/°C to-4.657pcm/ o C,-1.00518 pcm/ o C to 1.00649 pcm/ o C and-9.11361 pcm/ppm to-8.0751 pcm/ppm, respectively. For the water density reactivity coefficients, the positive reactivity occurs at the water temperature less than 190 o C. The calculation results show that the reactivity coefficients are accurate because the results have a very good agreement with the design value.
Core conversion design study of TRIGA Mark 2000 Bandung using MTR plate type fuel element
International Journal of Nuclear Energy Science and Technology
Science and Technology of Nuclear Installations
The in-house coupled neutronic and thermal-hydraulic (N/T-H) code of BATAN (National Nuclear Ener... more The in-house coupled neutronic and thermal-hydraulic (N/T-H) code of BATAN (National Nuclear Energy Agency of Indonesia), NODAL3, based on the few-group neutron diffusion equation in 3-dimensional geometry using the polynomial nodal method, has been verified with static and transient PWR benchmark cases. This paper reports the verification of NODAL3 code in the NEA-NSC PWR uncontrolled control rods withdrawal at zero power benchmark. The objective of this paper is to determine the accuracy of NODAL3 code in solving the continuously slow and fast reactivity insertions due to single and group of control rod bank withdrawn while the power and temperature increment are limited by the Doppler coefficient. The benchmark is chosen since many organizations participated using various methods and approximations, so the calculation results of NODAL3 can be compared to other codes’ results. The calculated parameters are performed for the steady-state, transient core averaged, and transient hot ...
Alternative Fueling Scheme for the Indonesian Experimental Power Reactor (10 MWth Pebble-Bed HTGR)
Energy Procedia
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA
The 10 MW HTR Indonesian Experimental Power Reactor (RDE reactor) is designed identical with the ... more The 10 MW HTR Indonesian Experimental Power Reactor (RDE reactor) is designed identical with the HTR-10 in China, conceptually. However, the review results showed that the spent fuel cask model which is used between two reactors is fully different, such as size and capacity. The proposed cask model in RDE reactor can hold 15 times more fuel pebbles than HTR-10 has. This research activities deal with the subcriticality analysis for the spent fuel cask of RDE reactor if using the HTR-10 cask model. The subcriticality condition is designed to meet the limit of safety value. The objective of this research is to determine the subcriticality value in the normal and accident events for the spent fuel cask when it is in the reactor building and the spent fuel cask room. All calculations were carried out by MCNP6.1 code. The selected external events are the water ingress (reactor room), water flood and the combination event of water flood and earthquake. The calculation results showed...
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA
NEUTRONICS ANALYSIS ON MINI TEST FUEL IN THE RSG-GAS CORE. Research of UMo fuel for research reac... more NEUTRONICS ANALYSIS ON MINI TEST FUEL IN THE RSG-GAS CORE. Research of UMo fuel for research reactor has been developing right now. The fuel of research reactor used is uranium low enrichment with high density. For supporting the development of fuel, an assessment of mini fuel in the RSG-GAS core was performed. The mini fuel are U7Mo-Al and U6Zr-Al with densitis of 7.0gU/cc and 5.2 gU/cc, respectively. The size of both fuel are the same namely 630x70.75x1.30 mm were inserted to the 3 plates of dummy fuel. Before being irradiated in the core, a calculation for safety analysis from neutronics and thermohydrolics aspects were required. However, in this paper will discuss safety analysis of the U7Mo-Al and U6Zr-Al mini fuels from neutronic point of view. The calculation was done using WIMSD-5B and Batan-3DIFF code. The result showed that both of the mini fuels could be irradiated in the RSG-GAS core with burn up less than 70 % within 12 cycles of operation without over limiting the ...
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA
VALIDATION OF FULL CORE GEOMETRY MODEL OF THE NODAL3 CODE IN THE PWR TRANSIENT BENCHMARK PROBLEMS... more VALIDATION OF FULL CORE GEOMETRY MODEL OF THE NODAL3 CODE IN THE PWR TRANSIENT BENCHMARK PROBLEMS. The coupled neutronic and thermal-hydraulic (T/H) code, NODAL3 code, has been validated in some PWR static benchmark and the NEACRP PWR transient benchmark cases. However, the NODAL3 code have not yet validated in the transient benchmark cases of a control rod assembly (CR) ejection at peripheral core using a full core geometry model, the C1 and C2 cases. By this research work, the accuracy of the NODAL3 code for one CR ejection or the unsymmetrical group of CRs ejection case can be validated. The calculations by the NODAL3 code have been carried out by the adiabatic method (AM) and the improved quasistatic method (IQS). All calculated transient parameters by the NODAL3 code were compared with the reference results by the PANTHER code. The maximum relative difference of 16% occurs in the calculated time of power maximum parameter by using the IQS method, while the relative difference ...
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA
The RSG-GAS reactor has a facility for irradiation of the fuel pin of nuclear power reactor, name... more The RSG-GAS reactor has a facility for irradiation of the fuel pin of nuclear power reactor, namely Power Ramp Test Facility (PRTF). The in-house fabrication PWR fuel pin has prepared for irradiations in the PRTF facility, currently, while the various enrichments of uranium are analyzed using the analytical tool. In the next step, it is planned to perform an irradiation of PHWR fuel pin sample of natural UO2 in the facility. Before irradiation in the core, it should be analyzed by using the analytical tool. The objectives of this paper are to optimize irradiation time based on the burn-up, the generated linear power and the neutron flux level at the target. The 3-dimension calculations have been carried out by using the CITATION code in the SRAC2006 code system. Since the coolant of the reactor is H2O, the effect of moderators in the pressurized tube, H2O and D2O, were analyzed, as well as pellet radius and moderator densities. The calculation results show that the higher linear pow...
International Journal of Engineering Research and, 2016
THE FEASIBILITY STUDY to USE UMo-Al FUEL AT RSG-GAS CORE. At present, The RSG-GAS is using U3Si2-... more THE FEASIBILITY STUDY to USE UMo-Al FUEL AT RSG-GAS CORE. At present, The RSG-GAS is using U3Si2-Al dispersion fuels with uranium density of 2.96 gU/cc and the. silicide uranium fuels will not be used anymore for the future. To anticipate the usage of other fuels in the RSG-GAS core, UMo-Al fuels were chosen. The UMo-Al fuel has many advantages, such as, it can be used at higher density in the reactor core. There are high uranium densities in UMo-Al dispersion fuels up to 15 gU/cc with numerous contents of Mo. In this analysis, the UMo fuel is used with density of 3.55 gU/cc and the contents consists of UMo9wt%-Al, Mo8wt%-Al, Mo7wt%-A and Mo6wt%-Al, called U9Mo, U8Mo, U7Mo and U6Mo respectively. The neutronic parameters, such as, reactivity balances, k-eff and power peaking factors of UMo-Al fuel with higher density (3.55 g U/cc) for the existing typical working core calculation have been carried out. The UMo-Al core criticality data were calculated using 2 dimension diffusion code Batan-EQUIL. The UMo-Al fuel macroscopic crossection data as the output of cell calculation WIMSD-B5 (ENDFVII.8) had been used for the calculation. The core calculations were performed using 2 and 3 dimension diffusion codes. The good fuel for RSG-GAS can be U7Mo-Al with the density of 3.55 g U/cc. The maximum radial, axial power peaking factor of the fuel at 31 cm is 1.28 and 1.30 when all control rods down and up respectively. Those are also met with the safety criteria. Indeed, the fuel of U7Mo with 3.55 gU/cc could be applied for the RSG-GAS core and operated for 960 MWD cycle length without any part of core configuration changes.
Fuel element burnup measurements for the equilibrium LEU silicide RSG GAS (MPR-30) core under a new fuel management strategy
Annals of Nuclear Energy, 2016
Science and Technology of Nuclear Installations, 2016
This paper reports the results of sensitivity analysis of the multidimension, multigroup neutron ... more This paper reports the results of sensitivity analysis of the multidimension, multigroup neutron diffusion NODAL3 code for the NEACRP 3D LWR core transient benchmarks (PWR). The code input parameters covered in the sensitivity analysis are the radial and axial node sizes (the number of radial node per fuel assembly and the number of axial layers), heat conduction node size in the fuel pellet and cladding, and the maximum time step. The output parameters considered in this analysis followed the above-mentioned core transient benchmarks, that is, power peak, time of power peak, power, averaged Doppler temperature, maximum fuel centerline temperature, and coolant outlet temperature at the end of simulation (5 s). The sensitivity analysis results showed that the radial node size and maximum time step give a significant effect on the transient parameters, especially the time of power peak, for the HZP and HFP conditions. The number of ring divisions for fuel pellet and cladding gives neg...
Desain Teras Alternatif Untuk Reaktor Riset Inovatif (Rri) Dari Aspek Neutronik
Tri Dasa Mega Jurnal Teknologi Reaktor Nuklir, Mar 28, 2015
Manajemen Teras RSG.GAS Berbaiian Bakar Silisida 4,5 Dan 4,8 G U/CC
Jurnal Sains Dan Teknologi Nuklir Indonesia, Aug 31, 2005
Jurnal Iptek Nuklir Ganendra, Jul 22, 2012
Persamaan difusi neutron dengan metode nodal telah menjadi metode standar dalam perhitungan param... more Persamaan difusi neutron dengan metode nodal telah menjadi metode standar dalam perhitungan parameter neutronik teras reaktor daya air ringan, seperti reaktor air bertekanan (PWR) atau air mendidih (BWR), karena waktu komputasi yang cepat dan hasilnya akurat. Paket program NODAL3 telah dikembangkan untuk menyelesaikan persamaan difusi neutron dengan metode nodal polinom (PNM) dalam geometri 3-dimensi (3-D). Validasi hasil perhitungan NODAL3 untuk kasus statis teras benchmark reaktor PWR (Pressurized Water Reactor), seperti IAEA-2D, BIBLIS, KOEBERG dan IAEA-3D, disajikan dalam penelitian ini. Teras benchmark yang dipilih mewakili kasus 2-D dan 3-D dan mempunyai karaktristik yang berbeda, sehingga dapat menentukan keakuratan NODAL3 dari aspek neutronik yang luas. Parameter neutronik statis yang dihitung adalah faktor perlipatan efektif, keff, faktor puncak daya (FPD) dan profil distribusi FPD ke arah aksial. Dibandingkan dengan acuan, hasil perhitungan NODAL3 menunjukkan bahwa untuk nilai keff terdapat perbedaan maksimum sebesar 0,006% (Δk). Sedangkan untuk FPD radial dan FPD aksial maksimum, selisih maksimum dengan acuan masing-masing sebesar-0,006 dan 0,051. Untuk kasus 3-D, hasil perhitungan NODAL3 konsisten dengan hasil perhitungan paket program nodal tervalidasi PARCS dan NESTLE. Riset ini menunjukkan bahwa akurasi paket program NODAL3 dalam menghitung parameter neutronik teras benchmark reaktor PWR, baik kasus 2-D dan 3-D, menunjukkan hasil yang sangat memuaskan. Oleh karena itu, paket program NODAL3 siap untuk diaplikasikan dalam analisis neutronik reaktor PWR yang riil.
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Papers by Tagor Sembiring