Indonesia Conference Directory

Corresponding Author
Indah Rosidah Maemunah

Institutions
ITB

Abstract
There have been performed calculation about blanket module, using either solid, liquid, or molten salt material. The aim of the comparison is not only to find the effective configuration for assuring the tritium self-sufficiency condition, but also to predict the material damage rate and the Helium production rate. In the TBR, the liquid/molten salt breeding material resulted a high TBR (>1.15), whereas Iron material as First Wall (FW) component produced a higher Helium concentration, compared to Cr, Mo, W material.

Keywords
FW, helium, molten salt, TBR

Topic
Nuclear and Radiation Computation

Link: https://ifory.id/abstract/tc4nYNHAhxMG

Corresponding Author
Fitria Miftasani

Institutions
Institut Teknologi Bandung

Abstract
The use of ZrC as a substitute for SiC has been proposed to improve the performance of coated fuel particles. In previous studies, a neutronic analysis on HTGR 30 MWt using ZrC and SiC has been done. The advance research conducted to see the effect of the use of TRIZO on HTGR with different power. In the present study, we use a power variation of 50-100 MWt and change the reactor geometry by apply additional fuel layer in the axial and radial direction from HTTR geometry as a reference. Neutronic analysis of the reactor was investigated by calculating the k-eff and k-inf values ​​using ZrC and SiC layers for coated fuel particle. Neutronic calculations are performed using the SRAC code with the JENDL 4.0 nuclear data library.

Keywords
HTGR, ZrC, SiC, TRISO

Topic
Nuclear and Radiation Computation

Link: https://ifory.id/abstract/xEPLYtemDAZ4

Corresponding Author
Rasito Tursinah

Institutions
Nuclear Physics Laboratory, Bandung Institute of Technology, Jl. Ganeca 10, Bandung, Indonesia
Email: rasito20[at]gmail.com

Abstract
A beam shaping assembly (BSA) has been designed in the beam tubes of G.A Siwabessy reactor for BNCT application. The BSA is used to decrease thermal neutron, fast neutron, and gamma dose of neutron beam output. The design was carried out in a simulation using the Monte Carlo method with the PHITS computer code. Design of BSA was obtained with TiF3 materials of 100 cm as a neutron moderator and gamma filter, Al of 2 cm as a reflector, aperture and shield using Pb and LiF. This design was able to reduce fast neutron and thermal neutrons to 100000 times, gamma doses to 10000 times, while decreasing epithermal neutron to 10000 times. To meet BNCT requirements, the BSA design should be placed at a depth of 100 - 200 cm of the radial type of RSG-GAS beam tubes.

Keywords
BSA, beam tubes, RSG-GAS, BNCT

Topic
Nuclear and Radiation Computation

Link: https://ifory.id/abstract/bMrhAC6mcgQF

Corresponding Author
Fitria Miftasani

Institutions
Institut Teknologi Bandung

Abstract
In this study, we designed a 100 MWt HTGR using High-Temperature Engineering Test Reactor (HTTR) 30 MWt as reference. The ZrC layer applied to the TRISO coated fuel particles as a substitute from the SiC layer. Geometry of the reactor is changed by adding the additional layer in the axial direction. The change in geometry is expected to extend the life of the reactor, and the use of ZrC Triso Coated Fuel Particle (TRIZO) is expected to improve the performance of coated fuel particles. Neutronic analysis is performed by calculating k-eff and k-inf using SRAC code with the JENDL 4.0 nuclear data library.

Keywords
HTGR, ZrC, SiC, TRISO

Topic
Nuclear and Radiation Computation

Link: https://ifory.id/abstract/DLkhyr9WFQMb

Corresponding Author
Nining Yuningsih

Institutions
aPhysics Study Program
bNuclear Physics and biophysics Research Division, Laboratory,Institut Teknologi Bandung,
Jl. Ganesha no. 10 Bandung 40132, Gedung Fisika FMIPA ITB Indonesia


1)ninng.yuningsih[at]students.itb.ac.id
2)dirwanto[at]fi.itb.ac.id

Abstract
East Nusa Tenggara (NTT) is an area with many islands. low rainfall intensity makes NTT especially Sabu Raijua as one of the areas with minimal water, especially clean water. Likewise with electricity that has not been sufficient. High-Temperature Gas Reactor (HTGR) is a type of reactor that produces not only electricity but also could be used for cogeneration applications, such as desalination of seawater. In this research, calculation of reactor design performed by Standard Thermal Reactor Analysis Code (SRAC) code and using Japanese Evaluated Nuclear Data Library (JENDL) 4.0 as nuclear data library. Meanwhile, seawater desalination was analyzed using the Desalination Economic Evaluation Program (DEEP) code. The reactor was designed to produced 150 MWt power while seawater desalination used Multi-Effect Desalination (MED) method. As a result, this reactor design can meet electricity demand in the Sabu Raijua region. Also, seawater desalination yields 110000 cubic meters per day which are meet the needs of clean water.

Keywords
High Temperature Gas Reactor (HTGR), seawater desalination, Multi Effect Desalination (MED)

Topic
Nuclear and Radiation Computation

Link: https://ifory.id/abstract/Nvg8d6PUQXM7

Corresponding Author
Arya Adhyaksa Waskita

Institutions
Center for Nuclear Reactor Technology and Safety, BATAN Indonesia

Abstract
A fission product release analysis code, called LPROF-BATAN, is being developed as part of acquiring the design and safety analysis capability of a pebble bed reactor. A triso-based fueled which applied in pebble bed reactor design assures a very limited fission product release to the environment below the hazardous limit. A specific analysis code is needed to performed quantitative analysis of the fission product which released from the pebble fuel, contained in the primary system, and finally release to environment. This paper presents the development LPROF-BATAN for the fission product release from the pebble fuel and contained in the primary system of the reactor. Comparison of forward Euler, backward Euler and Crank-Nicolson method in solving the initial value problem model of the fission product release is given in this paper. Comparison of LPROF-BATAN results to the available report of HTR-10 reactor design shows a good agreement.

Keywords
fission product release, initial value problem, triso-based fuel, pebble bed reactor

Topic
Nuclear and Radiation Computation

Link: https://ifory.id/abstract/F9MnNHQECJWj

Corresponding Author
Niken Rara Galih

Institutions
Laboratorium Fisika Nuklir,
Kelompok Keilmuan Fisika Nuklir dan Biofisika,
Fakultas Matematika dan Ilmu Pengetahuan Alam, Institut Teknologi Bandung,
Jl. Ganesha no. 10 Bandung, Indonesia, 40132

Abstract
As we know, human need of energy increase time to time. In order to fulfill human necessity of energy, in unison with the urge to provide long term clean energy for earth sustainability, building Nuclear Power Plant (NPP) seems to be the right answer. In the effort to support non-proliferation treaty, scientist had been developing fast reactor burning scheme design CANDLE (Sekimoto, 2010) which uses natural Uranium as its fuel. The Modified CANDLE (MCANDLE) as the modification of the previous CANDLE scheme has become the object of this research. Modified CANDLE divide the burning core into several discreet regions. This research uses Natural Uranium-Carbide and Natural Thorium-Carbide as fuel and Helium gas as coolant which was applied to an axial MCANDLE scheme reactor. The core size has been varied to optimize the design. Neutronic analysis has been applied to this research with burn up level, effective multiplication factor (k eff), infinite multiplication factor (k inf), and conversion ratio (CR) as observed parameter. Uranium percentage in fuel has been varied to reduce undesired power peaking. Neutronic calculation has been calculated using SRAC while the core design was calculated using FI-ITB-CH1 program. Combination of Uranium Carbide and Uranium Carbide – Thorium Carbide mixture fuel has reduced the power peaking factor of the reactor. Optimization has been reached as the radial length of the reactor set to 180cm and the axial length to 303cm.

Keywords
Modified CANDLE, burn up level, k eff, k inf, conversion ratio, power peaking

Topic
Nuclear and Radiation Computation

Link: https://ifory.id/abstract/xNwP8JzMQ3L9

Corresponding Author
Zaki Suud

Institutions
1 Nuclear and Biophysics Research Div. ITB
2 Dept of Physics Lampung Univiversity
3 Dept of Physics Jember University

Abstract
UTOP Accident is among important hypothetical accident in Pb-Bi Cooled Fast Reactors. In order to investigate the inherent safety performance of a Pb-Bi Cooled fast reactor a computer code to analyze Unprotected rod run-out Transient Overpower Accident (UTOP accident) is necessary. In this study improvement and parallelization of UTOP accident analyze code for Pb-Bi cooled fast reactors has been performed. The code adopts coupling of space time kinetic and transient thermal hydraulic analysis. Adiabatic approach of space time kinetic is adopted. The mathematical equations are discreetized and implemented on cluster computers using fortran language and MPI approach. The main challenge is the parallelization of the relatively tight coupled part of the program. The space time kinetic is basically relatively tightly coupled part so that difficult to be parallelized. On the other hands multi channels analysis in the thermal hydraulic part is particularly simpler for parallelization process. In general the program can significantly accelerated in the cluster computer with up to 40 core

Keywords
UTOP accident, inherent safety, MPI, space time kinetic

Topic
Nuclear and Radiation Computation

Link: https://ifory.id/abstract/DXhFWmJENanG

Corresponding Author
Nina Widiawati

Institutions
1Nuclear and Biophysics Department, Institut Teknologi Bandung, Indonesia.
2Department of Nuclear Safety Engineering, Tokyo, Tokyo City University, Japan
3Emeritus professor, Tokyo, Tokyo Institute of Technology

Abstract
Liquid metal cooled fast reactor is a reactor that included in the Generation IV reactor. Some liquid metal that has been used as fast reactor coolant are sodium (Na), Pb, and Pb-Bi. Based on several studies, Pb-based coolant shows better neutronic performance than Na. Currently, there are many studies related to the utilization of Pb208, which is one of the Pb stable isotopes as a coolant material. It is due to Pb208 has the lowest neutron absorption cross-section value among other stable Pb isotopes even compared to Pbnatural. Using pb208 isotope as a coolant can increase the k-eff value compared to natural Pb. In this research, a neutronic performance comparison between Pbnat and Pb208 coolant will be performed in a fast reactor with modified CANDLE burnup scheme. A reactor with a modified CANDLE burnup scheme could directly use natural uranium as fuel. The neutronic calculation performed using the SRAC2006 program. The utilization of liquid Pb208 as coolant can achieved the highest k-eff value among all coolant materials. Hence, the used of excellent coolant material in a reactor with a modified CANDLE burnup scheme is expected to has a superior neutron economy.

Keywords
Pbnat, Pbnat-Bi, Pb208, SRAC2006, Neutron economy, Coolant material, modified CANDLE.

Topic
Nuclear and Radiation Computation

Link: https://ifory.id/abstract/cpqYKnj4ayu2

Corresponding Author
Andrey Kosasih

Institutions
(a) Department of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Jl. Ganesa 10, Bandung 40132, Indonesia
(b) Nuclear Physics & Biophysics Research Division, Department of Physics, Faculty of
Mathematics and Natural Sciences, Bandung Institute of Technology, Jl. Ganesa 10, Bandung 40132, Indonesia
*awaris[at]fi.itb.ac.id

Abstract
The high temperature engineering test reactor (HTTR) is a block-type high-temperature gas-cooled reactor (HTGR) developed by Japan. This reactor is one of the Generation IV nuclear energy systems and can operate with coolant outlet temperature of 950°C. In this study, the neutronic analysis is carried out for the HTTR reactor with thorium fuel and helium coolant. As thorium has no naturally occurring fissile isotope, it requires other fissile isotope to sustain the nuclear chain reaction. In this study, U-233 is used as the fissile isotope. The fuel blocks used in the core vary from 3.3% to 7.5% of U-233 content. Several neutronic parameters are analyzed, such as effective multiplication factor, conversion ratio, neutron spectrum, power density distribution, and power peaking factor. The calculations are performed by PIJ and CITATION modules on SRAC2006 code with JENDL-4.0 as the nuclear data library. The cell-burnup calculations are conducted with two models, with and without microscopic cell definition in the fuel compact. The core calculations are conducted with triangular-z and hexagonal-z core geometry.

Keywords
HTTR, JENDL-4.0, SRAC2006, Thorium

Topic
Nuclear and Radiation Computation

Link: https://ifory.id/abstract/jYhXwxfdzKkT

Corresponding Author
Helen Raflis

Institutions
(a) Nuclear Physics and Biophysics Research Division, Physics Department, Faculty of Mathematics and Natural Science, Bandung Institute of Technology Indonesia, Jalan Ganesha 10 Bandung 40132, INDONESIA

*helenraflis[at]students.itb.ac.id

Abstract
The neutronics analysis of core configuration and dimension variation for modular Gas-cooled Fast Reactor (GFR) that selected by the generation IV international forum as one of six advanced reactor concepts has been done to understand of GFR performance. The main advantages of modular GFR compare another advanced reactor concept is using helium gas as main coolant due to figure out of feasibility of GFR core design need the simulation and modeling. In this paper, the variation of core configuration and dimension for core design have applied in radial, axial and radial-axial direction. The Monte Carlo method code named MCNP6 and OpenMC have been used for the criticality and isotope evaluation of design core GFR. The Monte Carlo method code provides the exact solution to solve the neutron transport equation in full-scale and heterogeneous three-dimensional (3D) geometry modeling using Evaluated Nuclear Data File (ENDF/B-VII.b5) nuclear data and continuous energy. The neutronics parameters characterized are the value of keff, power and neutron flux distribution, and burn-up level to know of the performance of GFR core design. The result of analysis showed that the core configuration in radial direction and dimension variation give the good understanding about the feasibility of GFR core design. In summary, by varied the core configuration in radial direction is prospective design for the next research.

Keywords
Neutronic Analysis, Modular GFR, Monte Carlo Method, Core Configuration, Core Dimension

Topic
Nuclear and Radiation Computation

Link: https://ifory.id/abstract/vkVT84mLGzhw

Corresponding Author
Feriska Handayani Irka

Institutions
(1)Nuclear Research group, FMIPA, ITB, Jl. Ganesha 10, Bandung, Indonesia
(2)Tokyo Institut Of Technology
*feriska.irka[at]gmail.com

Abstract
Gas Cooled Fast Reactor-GFR is one of the Generation IV reactors, a high temperature helium-cooled with a closed fuel cycle. Due to target operation on 2022-2030, this reactor type still need for further research and development technologies. In this paper, we investigated neutronics performances of GFR balance type core with modified CANDLE burn up scheme in radial direction. The power range of 300-700 MWTh. Modified CANDLE burn up scheme was chosen so that the reactor could operate with natural uranium input only. The neutronics calculation were performed using SRAC 2002 with JENDL 4.0 nuclear data library. The results show reactor could operate critically for 10 years without refueling with burn up level 20% HM

Keywords
Modified CANDLE in radial direction, output power, GFR, natural uranium, Generation IV

Topic
Nuclear and Radiation Computation

Link: https://ifory.id/abstract/VFrnR2dExZ83

Corresponding Author
Wahid Luthfi

Institutions
1) Centre for Nuclear Reactor Technology and Safety - National Nuclear Energy Agency, 80th Building of Center for Science and Technology, South Tangerang, Indonesia 15310

Abstract
Further study to simplify the modeling of pebble bed high temperature reactor core (HTR) has been widely developed before. From last calculation on actual fueled pebble to dummy ratio, 57:43, some variation of TRISO unit and Pebble unit is modelled to achieve its first criticality configuration. In this paper, some model that use 27000 pebble with 57:43 ratio and 100% fueled pebble is created to be used on burnup calculation se we could compare its K-EFF and nuclide inventory of it. From this burnup calculation, we could see that SC TRISO unit gives us faster calculation time followed by HCP TRISO unit and then FCC TRISO unit. On the other hand, BCC pebble unit had some consistent deviation from other pebble unit, and we need more study to know 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 reactor. On 57:43 ratio, absolute average deviation of K-EFF on burnup calculation is lower than 2% and 10% for nuclide inventory (mass). On 100% fueled pebble, its below 0.15% on K-EFF absolute deviation and below 8% on nuclide inventory deviation.

Keywords
TRISO, Pebble, Modelling, MCNPX, K-EFF, Burnup

Topic
Nuclear and Radiation Computation

Link: https://ifory.id/abstract/2qANtfXgFwWR

Corresponding Author
Zaki Suud

Institutions
1Nuclear and Biophysics Research Divisions, Bandung Institute of Technology
2National Nuclear Energy Agency, Indonesia

Abstract
Global Warming become more important issue in the recent decade. Related to this issue, energy conversion efficiency play important role in achieving economical design of NPP and directly influences thermal pollution to the environment as for the case of gas fueled power plants which becomes very competitive with advanced high temperature gas turbine system. In this study the feasibility to develop high temperature modular gas cooled fast reactors have been investigated. In this presentation the focus is in preliminary assessment of depressurized accident case/ High Temperature Ceramic materials are intensively used for structural materials. In this analysis assessment to estimate maximum temperature in the core during depressurized accident has been performed by employing radiation as the final mechanism to ttransfer the heat from the center of the core to the outer part. The natural circulation of air is assumed to be the final heat removal mechanism from the outer part of the core. Two dimensional R-Z geomentry calculation model has been employed to simulate the heat removal mechanism during depressurized accident. The results shows that by proper adjustment of the material and core design parameter it can be obtained designs which can survive depressurized accident inherently.

Keywords
GCFR, high temperature, energy conversion efficiency, depressurized accident, radiation

Topic
Nuclear and Radiation Computation

Link: https://ifory.id/abstract/WdJCLHnP3zNv

Corresponding Author
Cici Wulandari

Institutions
1Department of Physics, Faculty of Mathematics and Natural Sciences,
Institut Teknologi Bandung,INDONESIA
2Department of Physics and Department of Nuclear Science & Engineering,
Faculty of Mathematics and Natural Sciences,
Institut Teknologi Bandung,INDONESIA
*awaris[at]fi.itb.ac.id

Abstract
A preliminary conceptual design is conducted for 100 MW(electric) of Molten Salt Reactor (MSR) fueled with 233U. Neutronic parameters analysis of 100 MWe MSR are carried out by using the Monte Carlo method with MCNP6 program. The reactor can be operated for about 5 years without refueling and any exchange of graphite moderator. The unique liquid salt fuel form increases the reactor proliferation resistance because all radioactive isotopes in the core are hard to be separated. This one is a characteristic of Generation IV nuclear power system. The chemical compositions of fuel salt are LiF-BeF2-ThF4-233UF4. The Thorium utilization in the fuel provided sustainable energy in the reactor due to the breeding capability. The 233U concentration was varied to get the reactor criticality changes. The result showed that 233U loaded in fuel concentration is about 0.9 %mol for burnup 6.33GWd/MTU. This calculation is expected to be a preliminary analysis of the MSR development for future energy.

Keywords
MCNP6, Monte Carlo, MSR, Thorium, Uranium

Topic
Nuclear and Radiation Computation

Link: https://ifory.id/abstract/nutNH9zW4rAx

Corresponding Author
Dora Andris

Institutions
Nuclear Physics Laboratory
Bandung Institute of Technology
Jl Ganesha 10, Bandung, Indonesia

Abstract
Preliminary study on Modified CANLDE burnup strategy utilization in a helium cooled fast reactor have been performed. The Modified CANDLE burnup strategy is applied so that the reactor can operate using natural uranium fuel. Modified CANDLE burn-up strategy is applied in the radial direction because it is technically easier to implement. In modified CANDLE burnup strategy in radial direction, the active core is subdivided into ten regions with the same volume in the radial direction. When startup, each region contains fuel with different composition. The first region contains natural uranium (fresh fuel). The second region contains fuel from natural uranium burning for 10 years, the third region contains fuel from natural uranium burning for 20 years and so on. In this study, the Modified CANDLE burnup strategy in radial direction was designed using the MCNPX program. The fuel to be used is obtained from the calculation results of 2006 SRAC code system and FI ITB CH1 program. In the process, fuel will be shifted from one region to another until all regions are filled with new fuel. The result of calculation shows the reactor is in critical condition.

Keywords
HCFR; Modified CANDLE; radial direction

Topic
Nuclear and Radiation Computation

Link: https://ifory.id/abstract/KaeXzPBtbpkL

Corresponding Author
Feriska Handayani Irka

Institutions
(1)Nuclear Research group, FMIPA, ITB, Jl. Ganesha 10, Bandung, Indonesia
(2)Tokyo Institut Of Technology
*feriska.irka[at]gmail.com

Abstract
Gas cooled fast reactor-GFR is one of six types of generation IV reactors. The long terms goals of the Generation IV reactor is to develop a safe, proliferation-resistant nuclear system. Minimizing on enrichment of U-235 is one way to avoid the proliferation issue. Now, we even use directly natural uranium as the fuel in reactor core with applying Modified CANDLE burn up scheme. In this study variations of fuel region movement with modified CANDLE burn up scheme in radial direction of GFR was conducted. There are three variation of fuel region movement scheme that used in this study. The power range 300-550 MWTh. The neutronics calculation was performed by SRAC (PIJ-Citation) module. The result of calculation show that one of the movement scheme provides a critically value with 0,3% excess reactivity for begining of life (BOL) conditions.

Keywords
Modified CANDLE in radial direction, GFR, natural uranium, region movement, Generation IV

Topic
Nuclear and Radiation Computation

Link: https://ifory.id/abstract/jWNQJaCTb3mY

Corresponding Author
Indah Rosidah Maemunah

Institutions
ITB

Abstract
This paper addresses the neutronic analysis in Helium cooled blanket. The blanket design is justified in ITER where illustrated in solid breeder blanket, so the breeder material considered is ceramic solid breeding material that will have to be tritium self-sufficient condition. The Li2O, LiAlO2, Li2ZrO3, Li4SiO3 and Li2TiO3 are candidate of ceramic solid breeder to be tritium breeder in fusion blanket systems. The blanket had to be designed by beryllium as neutron multiplier because its property minimized these neutron losses. The work presented using the Monte Carlo simulation MCNP that developed in Los Alamos National Laboratory (2010) to evaluate the tritium breeding capability. It investigated neutron thermalization and multiplier, tritium production methods, and tritium breeding requirement.

Keywords
Breeding, ceramic solid, fusion, Helium, tritium

Topic
Nuclear and Radiation Computation

Link: https://ifory.id/abstract/mPyTt3uxMfq4

Corresponding Author
R. Andika Putra Dwijayanto

Institutions
a) Centre for Nuclear Reactor Technology and Safety - BATAN
Puspiptek Complex, OB No. 80, Serpong, South Tangerang 15310, Indonesia

b) Nuclear Physics and Biophysics Research Group, Department of Physics
Bandung Institute of Technology (ITB), Jl. Ganesha No. 10, Bandung 40132, Indonesia

*putra-dwijayanto[at]batan.go.id

Abstract
The existence of Tl-208 in thorium fuel cycle is a double-edged sword. Tl-208 is a strong 2.6 MeV gamma emitter, which acts as an effective proliferation barrier while in the same time complicating the spent fuel handling. To ensure the safety of the latter, the buildup of Tl-208 and its parent isotope, U-232, are necessary to be understood. This paper attempts to analyze U-232 and Tl-208 buildup in the Experimental Power Reactor (Reaktor Daya Eksperimental, RDE) fuel based on thorium cycle, using various U-233 isotopic vectors simulated with ORIGEN2.1 code. The simulation result shows that U-232-contaminated fresh fuels ended up with larger U-232 and Tl-208 activities at the end of cycle (EOC) compared with uncontaminated fresh fuel. However, their U-232 build-up rate are lower and even negative at one case. Then, lower U-233 purity resulted in larger U-232 and Tl-208 activities at EOC. This result implies a considerable difference of isotope buildup between the various U-233 vectors. Consequently, the thorium cycle-based RDE spent fuel handling should consider the isotopic vector of U-233 used in fresh fuel.

Keywords
RDE, thorium, Tl-208, U-232, ORIGEN2.1

Topic
Nuclear and Radiation Computation

Link: https://ifory.id/abstract/3hmyBuQzKXkC