Search

Analysis of the control rod insertion is important as it is closely related to reactor safety. Previously, the analysis has been carried out in RSG-GAS during static condition, not as a function of the fuel fraction. The RSG-GAS reactor in one cycle is a function of the fuel burn-up. It is necessary to analyze RSG-GAS core reactivity insertion as a function of the fuel burn-up to determine the behavior of the reactor, especially in uncontrolled operations such as continuous pulling of control rods. This analysis is carried out by the computer simulation method using WIMSD-5B and MTR-DYN codes, by observing power behavior as a function of time due to neutron chain reactions in the reactor core. Calculations are performed using point kinetics equation, and the feedback effect will be evaluated using static power coefficient and fuel burn-up function. Analyzes were performed for the core configuration of the core no. 99, by lifting the control rod or inserting positive reactivity to the core. The calculation results show that with the reactivity insertion of 0.5% Δk/k at start-up power of 1 W and 1 MW, safety limit is not exceeded either at the beginning, middle, or end of the cycle. The maximum temperature of the fuel is 135°C while the safety limit is 180°C. The margin from the safety limit is large, and therefore fuel damage is not possible when power excursion were to occur.

Liver cancer was the third leading cause of death from cancer in 2020 with 830,180 deaths worldwide. Radiotherapy is a common treatment method for liver cancer. Technological advances presented proton therapy and boron neutron capture therapy (BNCT) as alternatives with a lower dose on healthy organs. The objective of this research is to get a good dose distribution with higher tumor dose and lower healthy organ dose in proton therapy. A comparison with BNCT is done to get a better understanding of how both methods deliver the dose to treat the cancer while minimizing healthy organ doses. The research simulated proton therapy for cancer liver with Particle and Heavy Ions Transport Code System (PHITS), and a literature review for BNCT. The effectiveness of both methods were compared by tumor dose and liver dose. The optimal tumor dose for proton therapy is 86.01 Gy (W) with 0.67 Gy (W) liver dose. Proton therapy can replace conventional radiotherapy for tumors with complex shapes in dose delivery by utilizing its dose profile, while BNCT can give better tumor control on patients previously treated with conventional radiotherapy.

The power reactor with high-temperature gas-cooled reactor (HTGR) technology uses uranium as the reactor fuel. The energy from fission is converted to electrical energy or used for other needs such as hydrogen production or other research activities at high temperatures of around 700 °C. This operation does not allow the use of metal as the core material for the reactor. The material that fits the requirements as a core structure is graphite. Graphite material has specific characteristics, namely the parameters of the modulus of elasticity, coefficient of thermal expansion, and the volume which changes due to temperature and neutron dose. Because the structure of the reactor core is a vital component in the reactor, this research will develop a method for the design of the reactor core structure with graphite material. The design method is based on Design by Analysis which specifically refers to the strain analysis on each of the reactor core components. The design method developed is based on the finite element method. The object of this research is the side reflector made from the Toyo Tanso IG-110 series graphite. Based on the analysis of heat distribution and heat stress for the material before the effect of neutron exposure, the temperature distribution on the side reflector was found, as well as the displacement and heat stress that occurs. isotropic properties, Young's modulus and Poisson’s ratio values can be verified and estimated. The purpose of this research is to analyze the strain of the reactor core structure by taking into account the uncertainty of the graphite properties.

Crtical heat flux (CHF) is an important characteristic of nanofluids. The CHF measurements were carried out in nanofluid research at the Center for Applied Nuclear Science and Technology. These measurements are done manually using a variable power supply and a multimeter. However, it was difficult to record the voltage and current due to the sudden break of the wire. In this study, Arduino was used to measure CHF automatically. The voltage is applied to the wire and increases automatically along with the measurement of the voltage and current in the wire. The results of the voltage and current measurements were compared with a multimeter and were not significantly different. It can be concluded that the CHF measurement system using arduino can be used to measure nanofluid CHF.

The Indonesian Multipurpose Reactor, RSG-GAS reactor will accomplish its first lifetime in December 2020. The reactor has been operated in safe and reliable manner for about 33 years since it commenced in operation in 1987 to serve radioisotopes production, NAA, neutron beam experiments, material irradiation, and reactor physics experimental activities as well as training. The paper is intended to evaluate its in-core fuel management that is the conformance between the theory and implementation of the equilibrium core. Evaluation of the reactor operation parameter was carried out for core numbers 91 – 100. The data show that excess reactivity, shutdown reactivity and control rod reactivity have no significant difference at each core. The result shows that the BATAN-FUEL accurately determine the equilibrium core and its fuel loading pattern.This in-core fuel management of the RSG-GAS reactor supports the safety of reactor operation.

During the operation of the research reactor RSG-GAS, there are many design parameters should be verified based on postulated accidents. Several design basis accidents (DBA) such as loss of flow accident (LOFA) and reactivity-initiated accident (RIA) also have been conducted separately. This paper discusses about possibility of simultaneous accidents of LOFA and RIA. The accident analyses carry out calculation for transient condition during RIA, LOFA, and postulated accident of simultaneous LOFA-RIA. This study aims to conduct a safety analysis on simultaneous LOFA and RIA, and investigate the impact on safety margins. The calculations are conducted by using the PARET code. The maximum temperature of the center fuel meat at nominal power of 30 MW and steady state conditions is 126.10°C and MDNBR of 2.94. At transients condition, the maximum center fuel meat temperature for LOFA, RIA and simultaneous LOFA-RIA are consecutively 132.99°C, 135.67°C and 138.21°C, and the time of reactor trip are 3.2593s, 3.6494s and 2.7118s, respectively. While the MDNBR for LOFA, RIA and simultaneous LOFA-RIA are respectively at transient condition are 2.88, 2.58 and 2.63, respectively. It is shown that, simultaneous LOFA-RIA has the fastest trip time. In this case, the low flow trip occurs first in advance to over power trip. From these results, it can be concluded that the RSG-GAS has adequate safety margin against transient of simultaneous LOFA-RIA.Keywords: RSG-GAS, Simultaneous, LOFA, RIA, PARET

High Resolution Powder Diffractometer (HRPD) and Four Circle Diffractometer/Texture Diffractometer (FCD/TD) are two BATAN-owned neutron diffractometers which have been fully operational since 1992. These are used to investigate structure and texture of crystalline materials, respectively. Before analyzing, the acquired raw neutron diffraction data should first be processed in a specific way to achieve the suitable data format required by the analysis software. This data processing step is a repetitive task for every single experiment which is previously done manually and very time-consuming. The purpose of this development project was to optimize this step to be fully automatic and executable by a code. This work was performed by means of Python code utilizing the array manipulation in re-arranging and re-formatting the raw data. The resulted Python codes were named as hrpd.py and fcdtd.py. These have been successfully done and validated, making data processing step easier, simpler, and significantly faster with only 20 seconds or less required.Keywords: HRPD, FCD/TD, Automatic Data Processing, Neutron Diffraction, Python

This Section consists of Abstract Collection Page, Keywords Index Page, Author Index page, and Acknowledgment Page.

The mixed uranium-plutonium oxide fuel (MOX/UO2) is an interesting fuel for future power reactors. This is due to the large amount of plutonium that can be processed from spent fuel of nuclear plants or from plutonium weapons. MOX/UO2 fuel is very flexible to be applied in thermal reactors such as PWR and it is more economical than UO2 fuel. However, due to the different nature of neutron interactions of MOX in PWR, it will change the reactor core design parameters and also its safety characteristic. The purpose of this study is to determine the accuracy of SRAC2006 code system in generation of cross-sections and calculation of reactor core design parameters such as criticality, reactivity of control rods and radial power distribution. In this study, PWR MOX/UO2 Core Transient Benchmark is used to verify the code that models a MOX/UO2 fueled core. SRAC-CITATION result is different from DeCART by 0.339% from. SRAC-CITATION result of single rod worth in All Rods Out (ARO) conditions are quite good with a maximum difference of 6.34% compared to BARS code and 4.74% compared to PARCS code. In All Rods In (ARI) condition, SRAC-CITATION results compared to the PARCS code is quite good where the maximum difference is 9.72%, but compared to BARS code, it spikes up to 33.24% at maximum difference. In the other case, overall radial power density results are quite good compared to the reference. Its maximum deviation from DeCART code is 5.325% in ARO condition and 6.234% in ARI condition. Based on the results of these calculations, SRAC code system can be used to generate cross-section and to calculate some neutronic parameters. Hence, it can be used to evaluate the neutronic parameters of the MOX/UO2 PWR core design.Keywords: MOX/UO2 fuel, Criticality, Power peaking factor, SRAC2006

One of the barriers on the implementation of nuclear energy in Indonesia is public perception towards the safety of nuclear power plants (NPPs). Therefore, it is necessary to perform a study about the radiation impact of normal and abnormal operations of an NPP. In accordance to the program of Ministry of Research and Technology period 2020-2024, concerning the plan to build a small modular reactor (SMR)-type NPP, a radiation safety study has been performed for the 100 MWe Pressurized Water Reactor (PWR-100MWe). Source term release of radioactive substances into the environment from PWR-100MWe is a starting point in the study of the radiological consequences of reactor operation. Therefore, this paper will examine the PWR-100MWe source term under normal and abnormal operating conditions, according to the design and the design basis accident (DBA). The initial trigger of the DBA is Lost of Coolant Accident (LOCA) such as Small LOCA and Large LOCA. Due to the limitations of available SMR data, the study of PWR-100MWe source term refers to the assumption of the release fraction of fission products per subsystem in a larger 1000MWe PWR. It is expected from this assumption that pessimistic source term will be obtained. The study begins with calculation of PWR-100MWe core inventory using ORIGEN2 code based on PWR-100MWe reactor parameters. Through the mechanistic source term model and PWR-1000MWe release parameters, source terms will be obtained for normal operation and abnormal conditions i.e. DBA. Normal source term is used to calculate the consequences of normal operation, which will be used for environmental monitoring and environmental safety analysis of the site. Whereas accident source term is the basis for calculating the radiological consequences of accidents used for SAR documents and nuclear preparedness.Keywords: SMR, PWR-100MWe, normal operation, source term, accident

The reactivity value of the Bandung TRIGA 2000 reactor core has decreased over time, so the power generated by the reactor is also getting smaller, despite the control rod position is fully withdrawn. Therefore, it is necessary to reshuffle and refuel the fuel element to increase the excess reactivity by considering the safety parameters, such as axial and radial power peaking factors, DNBR, dTsat, and temperature on the cladding and in the center of the fuel element. The analyzed reactor safety parameters are the number of fuel elements, which varied at 105, 110, and 115 elements, as well as power peaking factor, which varied at 1.55, 1.65, 1.75, 1.85, and 1.95. The calculations were done using MCNP and COOLOD-N2 programs. If DNBR ≈ 1.3 is determined as the safety limit for the operation of the Bandung TRIGA 2000 reactor, at PPF 1.95 (105, 110, and 115 fuel elements), it can be considered to operate the reactor at the power of 600-700 kW. However, at PPF of 1.75 (105, 110, and 115 fuel elements), the reactor can be operated at the power of 700-800 kW, and at PPF of 1.55 (105, 110, and 115 fuel elements), the reactor can be considered for operation at the power of 800-900 kW. The results of these calculations can be used for consideration in determining the operating limits of the Bandung TRIGA 2000 reactor.Keywords: TRIGA 2000, fuel element, power peaking factor, DNBR, boiling

This Section consists of Abstract Collection Page, Keywords Index Page and Acknowledgment Page

Bandung TRIGA2000 Reactor, a General Atomic (GA)-made research reactor used for training, research andiIsotope production, has been upgraded to operate at power of 2000 kW using TRIGA fuel rod type. Recently, the TRIGA reactor fuel element producers are going to discontinue the production of TRIGA fuel element. To overcome the unavailability of TRIGA fuel element, BATAN planned to modify TRIGA2000 fuel type from rod-type to U3Si2-Al plate-type fuel with 19.75% enrichment, similar to the domestically fabricated one used in RSG-GAS. The carried out design emphasized on the determination of operation condition limits for setting the reactor protection system in accordance to the reactor safety calculation results. The conceptual design of the innovative fuel plate TRIGA reactor cooling system is expected to remove heat generated by fuels with nominal power of 1 MW up to 2 MW. The design is developed through modelling and safety analysis using COOLOD-N2 validated code. The safety margin is set to its flow instability at transient condition of the fuel plate, which is ≥ 2.38; departure from nucleate boiling ratio ≥1.50; and no onset of nucleate boiling, ΔTONB ≥ 0oC. The primary coolant flow rate accommodating the existing Bandung TRIGA reactor capability is as high as 50 kg/s. The analysis results show that at power of 1 MW, the reactor can safely operate, while at power of 2 MW the safety margin is exceeded. In other words, the plate TRIGA reactor that employs forced convection mode operates safely at 1 MW with excess power 120% of its nominal power.Keywords: 1 MW, Thermalhydraulic design, Steady state condition, TRIGA plate, Constant flowrate

The High Temperature Gas Cooled Reactor (HTGR) is considered as one of the nuclear reactors of generation-IV type in the future. The fuel handling system is one of the important processes in HTGR as well as in the design of Reaktor Daya Eksperimental (RDE). In the Fuel Handling System (FHS), the fuel pebble is transferred pneumatically along the pipe using carrier gas into the core of the reactor. Therefore, the pneumatic is an important system in operation stability of FHS. During the developing process of FHS-RDE, a branch pipe as a braking pipe system is provided on top of the pneumatic system to reduce the speed of the fuel discharged from the pneumatic pipe. The pneumatic pipe has an inner diameter of 65 mm and 20 m in length, whereas a branch pipe diameter for the braking system is 30 mm. The pneumatic system pressure is greater than the reactor cooling system pressure of 3.0 MPa. This work was performed to investigate the pressure drop and flow pattern of the braking system of FHS by various carrier gas inlet pressure. The analysis was carried out by Fluent 6.3 Software. Based on the design parameter of FHS used in the analysis, the results show that the performance of the braking system is not significant to reduce the pressure in the top region of the pneumatic pipe. To obtaining a significant reduction in pressure, and evaluation on the design of the branch pipe as well as the radius of curvature of the bend at the top pipe is suggested.Keywords: Pneumatic pipe, pressure, braking system, fuel handling of RDE, carrier gas

Molten salt reactor (MSR) is often associated with thorium fuel cycle, thanks to its excellent neutron economy and online reprocessing capability. However, since 233U, the fissile used in pure thorium fuel cycle, is not commercially available, the MSR must be started with other fissile nuclides. Different fissile yields different inherent safety characteristics, and thus must be assessed accordingly. This paper investigates the inherent safety aspects of one fluid MSR (OF-MSR) using various fissile fuel, namely low-enriched uranium (LEU), reactor grade plutonium (RGPu), and reactor grade plutonium + minor actinides (PuMA). The calculation was performed using MCNPX2.6.0 programme with ENDF/B-VII library. Parameters assessed are temperature coefficient of reactivity (TCR) and void coefficient of reactivity (VCR). The result shows that TCR for LEU, RGPu, and PuMA are -3.13 pcm, -2.02 pcm and -1.79 pcm, respectively. Meanwhile, the VCR is negative only for LEU, whilst RGPu and PuMA suffer from positive void reactivity. Therefore, for the OF-MSR design used in this study, LEU is the only safe option as OF-MSR starting fuel.Keywords: MSR, Temperature coefficient of reactivity, Void coefficient of reactivity, Low enriched uranium, Reactor grade plutonium, Minor actinides

This Section consists of Abstract Collection Page, Keywords Index Page and Acknowledgment Page

Cancer is a malignant tumor that destroys healthy cells. Cancer treatment can be done by several methods, one of which is BNCT. BNCT uses 10 B target which is injected into the human body, then it is irradiated with thermal or epithermal neutrons. Nuclear reaction will occur between boron and neutrons, producing alpha particle and lithium-7. The dose is estimated by how much boron and neutron should be given to the patient as a sum of number of boron, number of neutrons, number of protons, and number of gamma in the reaction of the boron and neutron. To calculate the dose, the authors simulated the reaction with Monte Carlo N Particle-X computer code. A water phantom was used to represent the human torso, as 75% of human body consists of water. Geometry designed in MCNPX is in cubic form containing water and a cancer cell with a radius of 2 cm. Neutron irradiation is simulated as originated from Kartini research reactor, modeled in cylindrical form to represent its aperture. The resulting total dose rate needed to destroy the cancer cell in GTV is 2.0814×10 14 Gy.s (76,38%) with an irradiation time of 1,4414×10 -13 s. In PTV the dose is 5.2295×10 13 Gy.s (19,19%) with irradiation time of 5.7367×10 -13 s. In CTV, required dose is 1.1866×10 13 Gy.s (4,35%) with an irradiation time of 2.5283×10 -12 s. In the water it is 1.9128×10 11 Gy.s (0,07%) with an irradiation time of 1,5684×10 -10 s. The irradiation time is extremely short since the modeling is based on water phantom instead of human body. Keywords: BNCT, Dose, Cancer, Water Phantom, MCNPX

The management of spent fuel is an issue of safety for Indonesia in the phase of designing RDE. Several studies regarding spent fuel are limited by geometrical characteristics and number of nuclides library. Therefore, different methodologies utilizing MCNPX2.6.0 were applied to get better information for further research. In this study, a single fuel pebble containing UO2, was burned using 5 cycles of multi-pass loading scheme for 1080 days to obtain the same energy as RDE’s core, which is about 79.90 GWd/MTU. The multiplication factor k-inf decreased at each cycle and stopped at 1.14575. The calculation results in the nuclides composition of the spent fuel after 1080 days of burning and 5 years of cooling containing 241 nuclides consist of 21 actinides and 220 nonactinides. Actinides with the highest activity of 8.96 Ci is with mass of 0.0867 g, whose half-life time is 14 years long. Nonactinides with the highest activity of 4.47 Ci is with mass of 0.0514 g, whose half-life time is 30.17 years long. The total activity of spent fuel pebble is 22.9 Ci with total mass of 5.28 g. The mass and activity data of each nuclide contained in the spent pebble will be used in the future research for performing safety analysis of the spent fuel storage tank.Keywords: Nuclides composition, Pebble, Spent fuel, RDE, MCNPX

The purpose of this study is to determine the value of corrosion rate influenced by coastal environment and seawater to nickel as a collimator base material for the application of boron neutron capture therapy (BNCT). In this research, the authors used 99.9% pure nickel as the reference material. Corrosion testing was carried out to determine the rate of corrosion of nickel as a base material for BNCT. After the specimens were formed, the test specimens were then corroded for 12 weeks, with various conditions such as indoor, outdoor environment, static seawater, and moving seawater. The results of this study indicated that in corrosion testing with indoor condition, the corrosion rate values are 0.61-1.00 mpy. For outdoor condition, the corrosion rate is 0.89-1.34 mpy. Meanwhile, at static seawater conditions, the corrosion rate is 0.97-1.24 mpy. Lastly, for moving seawater condition, the corrosion rate is 1.64-1.91 mpy. The results showed that corrosion resistance was relatively the same for all nickel exposed to corrosion in the coastal environment. Therefore, in regards to corrosion resistance, using nickel as a collimator base material for BNCT applications is considered as safe.Keywords: BNCT, Nickel, Corrosion, Coastal Environtment, Sea Water

As the effects of climate change are being felt all over the world, sustainability indicators such as water withdrawn per kilowatt-hour, are becoming more important in the decision-making process for large infrastructure projects. In Mongolia, we are deciding whether to use nuclear as a main power source. However, local droughts in Mongolia can be quite severe, occurring every 4-5 years and several countries have shown droughts to interrupt their power plant operations. This study collects data and conducts analyses to estimate sustainability indicators for a nuclear power plant life cycle and extends these analyses to understand how an event such as a drought would affect such indicators. The first part of this study is to provide background information regarding life cycle water use from power generation facilities. Our study focused on the APR-1400 nuclear power plant. If we account for drought frequency in Mongolia, the life cycle water withdrawal is estimated to be approximately 7,611 L/MWh for the nuclear power plant.Keywords: nuclear, sustainability, water, drought

Collection + Keywords Index + Acknowledgement

Public protection is one of important issues when operating nuclear facility. In case of accident occurs, the facility owner and related organizations shall make decision whether to evacuate people or not, based on the level of the accident and radiation dose rate released to the environment. In this study, as part of the decision support system for nuclear emergency response, a prototype of mobile radiation measurement system has been developed. The device consists of Geiger-Muller (GM)-based radiation measurement board, Global Positioning System (GPS) module, microcontroller board, and low power LoRa module for communication. Radiation dose rate along with its geoposition were recorded and sent to base station equipped with LoRa gateway for connecting LoRa network to TCP/IP-based network. The measurement data is then published to storage server using Message Queuing Telemetry Transport (MQTT) protocol. Power consumption, measurement of counter/timer accuracy, communication ranges testing, and radiation dose rate measurement were performed around Puspiptek area to demonstrate the functionality of the system.Keywords: Radiation monitoring, Decision Support System, Mobile, LoRa, GPS

Developing and licensing of digital Instrumentation and control (I&C) system for nuclear power plant (NPP) are challenging especially for the new construction since digital technology are composite with a very high complexity of many integrated systems. National Nuclear Energy Agency of Indonesia (BATAN), who design Reaktor Daya Eksperimental (RDE), should prepare the documents to meet the licensing requirements of national regulator in this case Nuclear Energy Regulatory Agency of Indonesia (BAPETEN). BAPETEN’s chairman regulation No.6 year of 2012 is the first national requirement which state requirements related to design of computer-based system concerning on safety of power reactor that should be followed. Since BAPETEN only denotes requirements without state which code and standards to be used, therefore BATAN can add references from International Nuclear Energy Agency (IAEA) guidelines. In this paper, requirement document traceability is developed to determine which code and standards should be used to verify and validate the I&C computer-based system of RDE. The hierarchy of regulatory and utility requirements are developed to guide the design basis documentation. Developing requirements analysis of computer-based I&C system RDE are completed after determining the design requirements from the utility and regulatory requirements. This methodology will help the design engineers to follow the utility requirements by concerning to the production, and follow the regulatory requirements concerning the safety aspect.Keywords: Computer-based system, I&C System, Requirements analysis, Licensing, RDE

In nulear emergency condition, after determining the initiating event and the type of the anomaly, operators should take counteractions to control the reactor to mitigate the accident and to bring back the plant to the safe condition. The actions should based on emergency operating procedures. In order to minimize the human error related to the actions, some necessary information is needed. Such kind of information is the consequence of the actions, which can be derived by modeling the counteractions. Multilevel flow modeling (MFM), a functional modeling, is chosen to model the counteraction with the consideration that it is based on cause-effect relations and consequence reasoning, it provides realization relationship which corresponds physical components with their functions, and it provides comprehensive diagnosis based on human perspective of the system objectives. The counteractions are represented by the control functions in the MFM. This paper discusses how to model the counteractions and the consequences of the actions to the system components, which are necessary to enhance situation awareness and to reduce human errors.Keywords: Operator actions, emergency operating procedures, multilevel flow modeling, control function, nuclear safety, human error

Boron Neutron Capture Therapy (BNCT) is one of the promising cancer therapy modalities due to its selectivity which only kills the cancer cells and does not damage healthy cells around cancer. In principle, BNCT utilizes the high ionization properties of alpha ( 4 He) and lithium ( 7 Li) particles derived from the reaction between epithermal and boron-10 neutrons ( 10 B + n → 7 Li + 4 He) in cells, where trace distance of alpha and lithium particles is equivalent with cell diameter. The neutron source used in BNCT can come from a reactor, as a condition for conducting BNCT therapy tests, there are five standard parameters that must be met for a neutron source to be used as a source, and the standards come from IAEA. This research is based on simulation using the MCNPX program which aims to optimize IAEA parameters that have been obtained in previous studies by changing the shape of the collimator geometry from cone shape to cylinder with variations diameter from 3, 5 and 10 cm and also the simulation divided into two schemes namely first moderator Al is placed in a position 9.5 cm behind the collimator and the second is the moderator Al is pressed into the base point of the aperture in the collimator. In this work, neutrons originated from Yogyakarta Kartini research reactor have the energy range in the continuous form. The results of the optimization on each scheme of the collimator are compared with the outputs that have been obtained in previous studies where the aperture of the collimator is in the cone shape. The most optimal output obtained from the results is a collimator with a diameter of 5 cm in the second scheme where the results of IAEA parameters that are produced (n/cm 2 s) = 2.18E+8, / (Gy-cm 2 /n) = 6.69E-13, / (Gy-cm 2 /n) = 2.44E-13, = 4.03E-01, and J / = 6.31E-01. These results can still be used for BNCT experiments but need a long irradiation time and when compared to previous studies, the output of the collimator with the diameter of 5 cm is more optimal. Keywords: BNCT, Collimator, IAEA Parameters, MCNPX, Cylindrical shape

G.A. Siwabessy Multipurpose Reactor (RSG-GAS) is a research reactor with thermal power of 30 MW located in the Serpong Nuclear Area (KNS), South Tangerang, Banten, Indonesia. Nuclear emergency preparedness of RSG-GAS needs to be improved by developing a decision support system for emergency response. This system covers three important aspects: accident source terms estimation, radioactive materials dispersion model into the atmosphere and radiological impact visualization. In this paper, radioactive materials dispersion during design basis accident (DBA) is modeled using HotSpot, by utilizing site-specific meteorological data. Based on the modelling, maximum effective dose and thyroid equivalent dose of 1.030 mSv and 26 mSv for the first 7 days of exposure are reached at distance of 1 km from the release point. These values are below IAEA generic criteria related to risk reduction of stochastic effects. The results of radioactive dispersion modeling and radiation dose calculations are integrated with Google Earth Pro to visualize radiological impact caused by a nuclear accident. Digital maps of demographic and land use data are overlayed on Google Earth Pro for more accurate impact estimation to take optimal emergency responses.Keywords: G.A. Siwabessy research reactor, Nuclear emergency, Atmospheric dispersion model, Decision support system, HotSpot codes

High Temperature Gas Cooled Reactor (HTGR) is a high temperature reactor type having nuclear fuels formed by small particles containing uranium in the core. One of HTGR designs is Pebble Bed Reactor (PBR), which utilizes helium gas flowing between pebble fuels in the core. The PBR is also the similar reactor being developed by Indonesia National Nuclear Energy Agency (BATAN) under the name of the Reaktor Daya Eksperimental (RDE) or Experimental Power Reactor (EPR) started in 2015. One important step of the EPR program is the completion of the detail design document of EPR, which should be submitted to the regulatory body at the end of 2018. The purpose of this research is to present preliminary results in the core temperature distribution in the EPR using the RELAP5/SCDAP/Mod3.4 to be complemented in the detail design document. Methodology of the calculation is by modelling the core section of the EPR design according to the determined procedures. The EPR core section consisting of the pebble bed, outlet channels, and hot gas plenum have been modelled to be simulated with 10 MWt. It shows that the core temperature distribution under assumed model of 4 core zones is below the limiting pebble temperature of 1,620 °C with the highest pebble temperature of 1,477.0 °C. The results are still preliminary and requires further researches by considering other factors such as more representative radial and axial power distribution, decrease of core mass flow, and heat loss to the reactor pressure vessel.Keywords: Pebble bed, core temperature, EPR, RELAP5 ANALISIS AWAL DISTRIBUSI TEMPERATUR TERAS REAKTOR DAYA EKSPERIMENTAL MENGGUNAKAN RELAP5. High Temperature Gas Cooled Reactor (HTGR) adalah reaktor tipe temperatur tinggi yang memiliki bahan bakar nukir dalam bentuk bola-bola kecil yang mengandung uranium. Salah satu desain HTGR adalah reaktor pebble bed (Pebble bed reactor/PBR) yang memanfaatkan gas helium sebagai pendingin yang mengalir di celah-celah bahan bakar bola di dalam teras. PBR juga merupakan tipe reaktor yang sedang dikembangkan oleh BATAN dengan nama reaktor daya eksperimental (RDE) yang dimulai pada 2015. Salah satu tahapan penting dalam program RDE adalah penyelesaian dokumen desain rinci yang harus dikirimkan ke badan pengawas pada akhir 2018. Tujuan penelitian adalah untuk menyajikan hasil-hasil awal pada distribusi temperatur di teras RDE menggunakan RELAP5/SCDAP/Mod3.4 sehingga dapat melengkapi isi dokumen desain rinci. Metode perhitungan adalah dengan memodelkan bagian teras RDE sesuai hasil penelitian sebelumnya. Bagian teras RDE yang dimodelkan terdiri dari pebble bed, kanal luaran, dan plenum gas bawah yang disimulasikan pada daya 10 MWt. Hasil simulasi menunjukkan bahwa distribusi temperatur teras dengan asumsi pembagian 4 zona teras mendapatkan temperatur tertinggi sebesar 1477 °C yang masih di bawah batasan temperatur di bola bahan bakar yaitu 1620 °C. Hasil yang diperoleh masih estimasi awal dan membutuhkan penelitian lebih lanjut dengan mempertimbangkan faktor-faktor lainnya seperti distribusi daya aksial dan radian yang lebih representatif, pengurangan aliran teras, dan kehilangan panas teras yang diserap oleh bejana reaktor.Kata kunci: Pebble bed, temperatur teras, RDE, RELAP5

Collection + Keywords Index + Authors Index + Acknowledgment

This section consists of Cover Page, Editorial Board Page, Peer Reviewer Page, Table of Content Page and Preface Page.

This section consists of Cover Page, Editorial Board Page, Peer Reviewer Page, Table of Content Page and Preface Page

This section consists of Cover Page, Editorial Board Page, Peer Reviewer Page, Table of Content Page and Preface Page

Due to TRIGA fuel elements are no longer produced by General Atomic, it is necessary to find a solution so that the Bandung TRIGA 2000 reactor can still be operated. One solution is to replace the type of fuel elements. Study on using the MTR plate type fuel elements as used in RSG-GAS Serpong has been done for the Bandung TRIGA 2000. Based on the results of the study using CFD computer program, it is found that Bandung TRIGA 2000 with plate type fuel elements cannot be operated up to 2000 kW power by natural convection cooling mode. Therefore, the reactor must be cooled by forced convection. The analysis using forced convection showed that for cooling flow rate of 50 kg/s and various temperatures of 35 o C, 35.5 o C and 36 o C, the surface temperature of the fuel element is between 110.37 o C and 111.27 o C. Meanwhile, the cooling water temperature in the corresponding position is between 61.03 o C and 61.95 o C. In this operation condition, the surface temperatures of fuel elements can approach the saturation temperature and nucleat boiling started to occur. Hence, the use of cooling flow rate entering core less than 50 kg/s should be avoided. The surface temperature of fuel elements decreased under saturation temperature if cooling flow rate is greater than 65 kg/s. The surface temperature of fuel elements is achieved at 96.65 o C and coolant temperature in the corresponding position was 54.38 o C . Keywords: Bandung research reactor, plate type fuel element, thermohydraulic, CFD code ANALISIS TERMOHIDROLIK TERAS REAKTOR RISET BANDUNG BERELEMEN BAKAR TIPE PELAT MENGGUNAKAN PROGRAM CFD. Mengingat tidak diproduksinya lagi elemen bakar TRIGA oleh General Atomic, maka perlu diusahakan suatu solusi agar reaktor TRIGA 2000 Bandung dapat tetap beroperasi. Salah satu solusi adalah dengan melakukan penggantian tipe elemen bakar. Pada studi ini telah dianalisis penggunaan elemen bakar tipe pelat yang sejenis dengan yang digunakan di RSG-GAS Serpong, untuk digunakankan pada teras reaktor TRIGA 2000 Bandung. Berdasarkan hasil penelitian yang telah dilakukan dengan menggunakan program komputer CFD, diketahui bahwa reaktor TRIGA berelemen bakar tipe pelat tidak dapat dioperasikan pada daya 2000 kW dengan menggunakan moda pendinginan konveksi alamiah seperti yang digunakan saat ini. Untuk kondisi ini, pendinginan dilakukan dengan moda pendinginan konveksi paksa. Hasil analisis konveksi paksa menunjukkan bahwa dengan menggunakan laju alir pendingin pompa 50 kg/s dan variasi temperatur pada 35 o C, 35,5 o C dan 36 o C, diperoleh temperatur permukaan pelat elemen bakar antara 110,37 o C – 111,27 o C dan temperatur pendinginnya pada posisi terkait antara 61,03 o C – 61,95 o C. Temperatur permukaan pelat elemen bakar ini mendekati temperatur saturasi dan tentunya telah mulai terjadi pendidihan inti, sehingga penggunaan laju alir pendingin masuk teras reaktor kurang dari 50 kg/s perlu dihindari. Temperatur permukaan pelat elemen bakar mulai menurun menjauhi temperatur saturasi jika digunakan laju alir pendingin lebih besar dari 65 kg/s, dengan temperatur permukaan pelat elemen bakar 96,65 o C dan temperatur pendinginnya pada posisi terkait 54,38 o C. Kata kunci: Reaktor riset Bandung, elemen bakar tipe pelat, termohidrolik, program CFD

Cover + Editorial Board + TOC + Preface

Nowadays, pumps are being widely used in the thermal power generation including nuclear power plants. Reaktor Daya Eksperimental (RDE) is a proposed nuclear reactor concept for the type of nuclear power plant in Indonesia. This RDE has thermal power 10 MW th , and uses a feedwater pump within its steam cycle. The performance of feedwater pump depends on size and geometry of impeller model, such as the number of blades and the blade angle. The purpose of this study is to perform a preliminary design on an impeller of feedwater pump for RDE and to simulate its performance characteristics. The Fortran code is used as an aid in data calculation in order to rapidly compute the blade shape of feedwater pump impeller, particularly for a RDE case. The calculations analyses is solved by utilizing empirical correlations, which are related to size and geometry of a pump impeller model, while performance characteristics analysis is done based on velocity triangle diagram. The effect of leakage, pass through the impeller due to the required clearances between the feedwater pump impeller and the volute channel, is also considered. Comparison between the feedwater pump of HTR-10 and of RDE shows similarity in the trend line of curve shape. These characteristics curves will be very useful for the values prediction of performance of a RDE feedwater pump. Preliminary design of feedwater pump provides the size and geometry of impeller blade model with 5-blades, inlet angle 14.5 degrees, exit angle 25 degrees, inside diameter 81.3 mm, exit diameter 275.2 mm, thickness 4.7 mm, and height 14.1 mm. In addition, the optimal values of performance characteristics were obtained when flow capacity was 4.8 kg/s, fluid head was 29.1 m, shaft mechanical power was 2.64 kW, and efficiency was 52 % at rotational speed 1750 rpm. Keywords: Blade, impeller, pump, RDE DESAIN AWAL IMPELER POMPA AIR UMPAN RDE. Saat ini, pompa digunakan secara luas dalam pembangkit tenaga termal termasuk pembangkit listrik tenaga nuklir. Reaktor Daya Eksperimental (RDE) merupakan konsep reaktor nuklir yang diusulkan untuk tipe PLTN di Indonesia. RDE ini memiliki daya termal 10 MW th , dan menggunakan pompa air umpan dalam siklus uapnya. Kinerja pompa air umpan bergantung pada ukuran dan geometri model impeller, seperti jumlah sudu dan sudut sudu. Tujuan dari penelitian ini adalah untuk membuat rancangan awal impeller pompa air umpan untuk RDE dan untuk mensimulasikan karakteristik kinerjanya. Kode Fortran digunakan sebagai bantuan dalam penghitungan data untuk untuk mengkalkulasi secara cepat bentuk sudu impeller pompa air umpan, terutama pada kasus RDE. Analisis perhitungan dipecahkan menggunakan korelasi empiris yang terkait dengan ukuran dan geometri model impeller pompa, sedangkan analisis karakteristik kinerja dilakukan berdasarkan diagram segitiga kecepatan. Pengaruh bocoran, melalui impeler akibat celah yang diperlukan antara impeller pompa air umpan dan saluran volute, juga dipertimbangkan. Perbandingan antara pompa air umpan HTR-10 dan RDE menunjukkan kemiripan dalam garis tren bentuk kurva. Kurva karakteristik ini akan sangat berguna untuk perkiraan nilai kinerja pompa air umpan RDE. Desain awal pompa air umpan memberikan ukuran dan geometri model sudu impeller dengan 5-sudu, sudut masuk 14,5 derajat, sudut keluar 25 derajat, diameter dalam 81,3 mm, diameter luar 275,2 mm, ketebalan 4,7 mm, dan tinggi 14,1 mm. Selain itu, nilai optimal karakteristik kinerja diperoleh ketika kapasitas aliran 4,8 kg/s, head fluida 29,1 m, tenaga mekanik poros 2,64 kW, dan efisiensi 52 % pada kecepatan putaran 1750 rpm. Kata kunci: Sudu, impeler, pompa, RDE

Radiation shield at Boron Neutron Capture Therapy (BNCT) facility based on D-D Neutron Generator 2.4 MeV has been modified with pre-designed beam shaping assembly (BSA). Modeling includes the material and thickness used in the radiation shield. This radiation shield is expected to protect workers from radiation doses rate that is not exceed 20 mSv·year-1 of dose limit values. The selected materials are barite, paraffin, polyethylene and lead. Calculations were performed using the MCNPX program with tally F4 to determine the dose rate coming out of the radiation shield not exceeding the radiation dose rate of 10 μSv·hr-1. Design 3 was chosen as the recommended model of the four models that have been made. The 3rd shield design uses a 100 cm thickness of barite concrete as primamary layer to surrounding 100 cm x 100 cm x 166.4 cm room, and a 40 cm borated polyethylene surrounding the barite concrete material. Then 10 cm barite concrete and 10 cm of borated polyethylene are added to reduce the primary radiation straight from the BSA after leaving the main layer. The largest dose rate was 4.58 μSv·h-1 on cell 227 and average radiation dose rate 0.65 μSv·hr-1. The dose rates are lower than the lethal dose that is allowed by BAPETEN for radiation worker lethal dose.Keywords: Radiation shield, tally, radiation dose rate, BSA, BNCT PEMODELAN PERISAI RADIASI PADA FASILITAS BORON NEUTRON CAPTURE THERAPY BERBASIS GENERATOR NEUTRON D-D 2,4 MeV. Telah dimodelkan perisai radiasi pada fasilitas Boron Neutron Capture Therapy (BNCT) berbasis reaksi D-D pada Neutron Generator 2,4 MeV dengan Beam Shaping Assembly (BSA) yang telah didesain sebelumnya. Pemodelan ini dilakukan untuk memperoleh suatu desain perisai radiasi untuk fasilitas BNCT berbasis generator neutron 2,4 MeV. Pemodelan dilakukan dengan cara memvariasikan bahan dan ketebalan perisasi radiasi. Bahan yang dipilih adalah beton barit, parafin, polietilen terborasi dan timbal. Perhitungan dilakukan menggunakan program MCNPX dengan tally F4 untuk menentukan laju dosis yang keluar dari perisai radiasi. Desain periasi radiasi dinyatakan optimal jika radiasi yang dihasilkan diluar perisai radiasi tidak melebihi Nilai Batas Dosis (NBD) yang telah ditentukan oleh BAPETEN. Hasilnya, diperoleh suatu desain perisai radiasi menggunakan lapisan utama beton barit setebal 100 cm yang mengelilingi ruangan 100 cm x 100 cm x 166,4 cm dan polietilen terborasi 40 cm yang mengelilingi bahan beton barit. Kemudian ditambahkan beton barit 10 cm dan polietilen terborasi 10 cm untuk mengurangi radiasi primer yang lurus dari BSA setelah keluar dari lapisan utama. Laju dosis terbesar adalah 4,58 μSv·jam-1 pada sel 227 dan laju dosis rata-rata yang dihasilkan adalah sebesar 0,65 µSv·jam-1. Nilai laju dosis tersebut masih dibawah ambang batas NBD yang diperbolehkan oleh BAPETEN untuk pekerja radiasi.Kata kunci: Perisai radiasi, tally, laju dosis radiasi, BSA, BNCT

A study to optimize a model of neutron radiation shielding for BNCT facility in the irradiation room has been performed. The collimator used in this study is a predesigned collimator from earlier studies. The model includes the selection of the materials and the thickness of materials used for radiation shield. The radiation shield is required to absorb leaking radiation in order to protect workers at the threshold dose of 20 mSv/year. The considered materials were barite concrete, paraffin, stainless steel 304 and lead. The leaking neutron radiation dose rates have been determined using Monte Carlo N Particle Version Extended (MCNPX) with a radiation dose limit rate that is less than 10 µSv/hour. This dose limit is in accordance with BAPETEN regulation related the threshold dose for workers, in which the working duration is 8 hours per day and 5 days per week. It is recommended that the best model for the irradiation room has a dimension 30 cm width, 30 cm length, 30 cm height and a main layer of irradiation room shielding made from the material paraffin which is 68 cm thickness on the left side and bottom of the irradiation room, 70 cm thickness on the right side of the iradiation room, 45 cm thickness on the front of the irradiation room and 67 cm thickness on the top of the irradiation room. The additional layers of 15 cm and 10 cm thickness are used along with paraffin in order to reduce the intensity of primary radiation from piercing the beamport after two primary layers. There is no neutron radiation leakage in this model.Keywords: Radiation shielding, BNCT, MCNPX, radiation dose rate, piercing beamport. OPTIMASI PERISAI RADIASI NEUTRON FASILITAS RUANGAN IRADIASI UNTUK BORON NEUTRON CAPTURE CANCER THERAPY (BNCT) DENGAN SUMBER BEAMPORT TEMBUS REAKTOR KARTINI. Telah dilakukan pemodelan perisai radiasi neutron untuk fasilitas Boron Neutron Capture Therapy (BNCT) pada sekeliling ruangan iradiasi. Pemodelan mencakup pemilihan bahan dan tebal yang digunakan untuk perisai radiasi. Perisai diharuskan mampu menahan radiasi yang keluar ruangan sehingga dosis radiasi berada di bawah ambang dosis bagi pekerja radiasi sebesar 20 mSv/tahun. Bahan yang dipertimbangkan adalah beton barit, paraffin, stainless steel 304 dan timbal. Perhitungan laju dosis neutron epitermal dilakukan dengan menggunakan program Monte Carlo N Particle Version Extended (MCNPX) dengan batasan laju dosis radiasi kurang dari 10 µSv/jam, sesuai dengan peraturan Kepala BAPETEN mengenai batas ambang laju dosis pekerja radiasi, dengan asumsi perhitungan waktu kerja 8 jam per hari dan 5 hari per minggu. Desain pertama dari empat desain yang telah dibuat kemudian dipilih sebagai desain yang direkomendasikan dengan laju dosis di bawah batas ambang 10 µSv/jam. Ruangan iradiasi memiliki dimensi panjang 30 cm, lebar 30 cm dan tinggi 30 cm. Lapisan utama perisai pada desain pertama berbahan paraffin setebal 68 cm pada sisi kiri dan bawah ruangan, 70 cm pada sisi kanan ruangan, 45 cm pada sisi depan ruangan dan 67 cm pada sisi atas ruangan. Paraffin setebal 15 cm dan 10 cm ditambahkan sebagai peredam intensitas radiasi primer dari beamport tembus yang masih cukup besar.Kata Kunci: perisai radiasi, BNCT, MCNPX, laju dosis radiasi, beamport tembus.

Cancer is a dangerous disease caused by the growth of a mass of cells that are unnatural and uncontrollable. Glioblastoma, also called as glioblastoma multiforme (GBM), is one of dangerous brain cancer. The dismal prognosis associated with glioblastoma is attributable not only to its aggressive and infiltrative behavior, but also to its location typically deep in the parenchyma of the brain. In resolving this chalenge, the BNCT method can be a solution. This study aims to calculate BNCT dosimetry in different of cancer positions and irradiation geometries using PHITS code. The results show that the deeper the cancers target at brain the slower the total absorbed dose rate of cancer target. It takes a longer treatment time. Based on the treatment time and total absorbed dose rate of cancer target, the TOP irradiation geometry is an appropriate choice in treating the cancer target in this case. To achieve the histopathological cure of GBM at the primary site, the absorbed dose of brain was calculated to be 1.07 Gy and 1.64 Gy for the LLAT and PA irradiation geometry, respectively. While, for cancer position of 3 cm, 5 cm, 7.15 cm, 9 cm, and 11 cm, the absorbed dose of brain is 0.25 Gy, 0.48 Gy, 0.85 Gy, 1.33 Gy, and 2.01 Gy, respectively. In addition to the stochastic effect, it was found also deterministic effects that may be produced such as cataracts. Keywords: BNCT dosimetry; GBM; brain cancer cases; PHITS; MIRD phantom PERHITUNGAN DOSIMETRI BNCT PADA KANKER OTAK BERBASIS REAKTOR RISET KARTINI MENGGUNAKAN PROGRAM PHITS. Kanker merupakansalahsatu penyakit berbahaya yang diakibatkan oleh tumbuhnya sekumpulan massa sel-sel yang tidak wajar dan tidak terkendali . Salah satu penyakit kanker otak yang berbahaya adalah Glioblastoma atau yang biasa disebut Glioblastoma Multiforme (GBM). Prognosis suram terkait dengan GBM tidak hanya untuk perilaku agresif dan infiltrasi, tetapi juga terhadaplokasi yang jauh di dalam parenkim otak. Untuk menjawab hal tersebut, Boron Neutron Capture Therapy (BNCT) dapat menjadi solusi. Penilitian ini bertujuan untuk menghitung dosimetri BNCT dalam berbagai posisikan kerdan geometri penyinaran dengan menggunakan program PHITS. Hasil perhitungan menunjukkan bahwa semakin dalam target kanker di otak maka semakin kecil total laju dosis serap dari target kanker. Semakin dalam target kanker di otak dibutuhkan waktu pengobatan yang semakin lama. Berdasarkan waktu pengobatan dan laju dosis serap dari target kanker, bidang penyinaran TOP merupakan pilihan yang tepat dalam mengobati target kanker dalam kasus ini. Untuk mencapai penyembuhan GBM secara histopatologis di lokasi utama, dosis serap dari otak dihitung berturut-turut sebesar 1,07 Gy dan 1,64 Gy untuk bidang penyinaran LLAT dan PA. Sedangkan, untuk posisi kanker 3 cm, 5 cm, 7,15 cm, 9 cm, dan 11 cm, berturut-turut dosis serap dari otak adalah 0,25 Gy, 0,48 Gy, 0,85 Gy, 1,33 Gy, and 2,01 Gy. Selain adanya efek stokastik, ditemukan juga efek deterministik yang mungkin dihasilkan seperti katarak. Kata kunci: Dosimetri BNCT, GBM, kasuskankerotak, geometripenyinaran, posisikanker, ORNLMIRD phantom.

One of the important things in reactor safety is the value of inherent safety parameter namely reactivity coefficient. These inherent safety parameters are fuel and moderator temperature coefficients of reactivity. The objective of the study is to obtain the change of those reactivity coefficients as a function of fuel burn up during the cycle operation of AP 1000 reactor core. Fuel and moderator temperature coefficients of reactivity and in addition moderator density coefficient of reactivity were calculated using SRAC 2006 and NODAL3 computer codes. Cross section generation of all core material was done by SRAC 2006 Code. The calculation of core reactivity as a function of temperature and burn up were carried out using NODAL3 Code. The results show that all reactivity coefficients of AP 1000 reactor core are always negative during the operation cycles and the values are in a good agreement to the design. It can be concluded that the AP 1000 core has a good inherent safety of its fuelKeywords: reactivity coefficient, burn up, AP1000, NODAL3. ANALISIS PERUBAHAN KOEFISIEN REAKTIVITAS AKIBAT FRAKSI BAKAR TERAS REAKTOR AP1000 MENGGUNAKAN NODAL3. Salah satu hal yang sangat penting dalam analisis kecelakaan pada reactor daya adalah koefisien reaktivitas untuk mengontrol daya reaktor. Penelitian ini bertujuan menentukan koefisien reaktivitas akibat perubahan fraksi bakar pada reaktor AP1000. Koefisien reaktivitas yang akan dihitung adalah koefisien reaktivitas bahan bakar dan moderator yang sering disebut inherent factor. Selain itu juga akan dihitung koefisien konsentrasi boron dan kerapatan moderator. Semua koefisien reaktivitas ini dihitung saat terjadi perubahan fraksi bakar untuk mempertimbangkan produk fisi dan konsumsi bahan bakar. Perhitungan neutronik teras reactor disimulasi dengan menggunakan program SRAC2006 dan NODAL3. Perhitungan tampang lintang seluruh perangkat bahan bakar dan batang kendali reaktor AP1000 dilakukan dengan program SRAC2006. Perhitungan parameter neutronik sebagai fungsi temperature dan fraksi bakar dilakukan menggunakan program NODAL3. Perhitungan koefisien reaktivitas ditentukan berdasarkan perbedaan nilai reaktivitas. Hasil perhitungan menunjukkan bahwa koefisien reaktivitas teras reaktor AP 1000 selalu berharga negative untuk sepanjang siklus operasinya dan mendekati harga desain. Kesimpulan yang dapat ditarik adalah bahwa teras AP 10000 mempunyai keselamatan melekat yang baik.Kata kunci: koefisien reaktivitas, fraksi bakar, AP 1000, NODAL3.

This article involves two main objectives of BNCT system. The first goal includes optimization of 30 MeV Cyclotron-based Boron Neutron Capture Therapy (BNCT) beam shaping assembly. The second goal is to calculate the neutron flux and dosimetry system of BNCT in the head and neck soft tissue sarcoma. A series of simulations has been carried out using a Monte Carlo N Particle X program to find out the final composition and configuration of a beam shaping assembly design to moderate the fast neutron flux, which is generated from the thick beryllium target. The final configuration of the beam shaping assembly design includes a 39 cm aluminum moderator, 8.2 cm of lithium fluoride as a fast neutron filter and a 0.5 cm boron carbide as a thermal neutron filter. Bismuth, lead fluoride, and lead were chosen as the aperture, reflector, and gamma shielding, respectively. Epithermal neutron fluxes in the suggested design were 2.83 x 10 9 n/s cm -2 , while other IAEA parameters for BNCT beam shaping assembly design have been satisfied. In the next step, its dosimetry for head and neck soft tissue sarcoma is simulated by varying the concentration of boron compounds in ORNL neck phantom model to obtain the optimal dosimetry results. MCNPX calculation showed that the optimal depth for thermal neutrons was 4.8 cm in tissue phantom with the maximum dose rate found in the GTV on each boron concentration variation. The irradiation time needed for this therapy were less than an hour for each level of boron concentration. Keywords: Optimization, Beam Shaping Assembly, BNCT, Dosimetry, 30 MeV Cyclotron, MCNPX. OPTIMASI DESAIN KOLIMATOR NEUTRON UNTUK SISTEM BNCT DAN UJI DOSIMETRINYA MENGGUNAKAN PROGRAM MCNPX . Telah dilakukan penelitian tentang sistem BNCT yang meliputi dua tahapan simulasi dengan menggunakan program MCNPX yaitu uji simulasi untuk optimasi desain kolimator neutron untuk sistem BNCT berbasis Siklotron 30 MeV dan uji simulasi untuk menghitung fluks neutron dan dosimetri radiasi pada kanker sarkoma jaringan lunak pada leher dan kepala. Tujuan simulasi untuk mendapatkan desain kolimator yang paling optimal dalam memoderasi fluks neutron cepat yang dihasilkan dari sistem target berilium sehingga dapat dihasilkan fluks neutron yang sesuai untuk sistem BNCT. Uji optimasi dilakukan dengan cara memvariasikan bahan dan ketebalan masing-masing komponen dalam kolimator seperi reflektor, moderator, filter neutron cepat, filter neutron thermal, filter radiasi gamma dan lubang keluaran. Desain kolimator yang diperoleh dari hasil optimasi tersusun atas moderator berbahan Al dengan ketebalan 39 cm, filter neutron cepat berbahan LiF 2 setebal 8,2 cm, dan filter neutron thermal berbahan B 4 C setebal 0,5 cm. Untuk reflektor, filter radiasi gamma dan lubang keluaran masing-masing menggunakan bahan PbF 2 , Pb dan Bi. Fluks neutron epithermal yang dihasilkan dari kolimator yang didesain adalah sebesar 2,83 x 10 9 n/s cm -2 dan telah memenuhi seluruh parameter fluks neutron yang sesuai untuk sistem BNCT. Selanjutnya uji simulasi dosimetri pada kanker sarkoma jaringan lunak pada leher dan kepala dilakukan dengan cara memvariasikan konsentrasi senyawa boron pada model phantom leher manusia (ORNL). Selanjutnya model phantom tersebut diiradiasi dengan fluks neutron yang berasal dari kolimator yang telah didesain sebelumnya. Hasilnya, fluks neutron thermal mencapai nilai tertinggi pada kedalaman 4,8 cm di dalam model phantom leher ORNL dengan laju dosis tertinggi terletak pada area jaringan kanker. Untuk masing-masing variasi konsentrasi senyawa boron pada model phantom leher ORNL supaya dapat mematikan jaringan kanker, membutukan waktu iradiasi neutron kurang dari satu jam. Kata kunci: Optimasi, Kolimator, BNCT, Dosimetri, Siklotron 30 MeV, MCNPX

Radiation dose assessment to determine the potential of radiological impacts of various installations within nuclear facility complex is necessary to ensure evironmental and public safety. A simple generic model-based method for calculating radiation doses caused by the release of radioactive substances into the environment has been published by the International Atomic Energy Agency (IAEA) as the Safety Report Series No. 19 (SRS-19). In order to assist the application of the assessment method and a basis for the development of more complex assessment methods, an open-source based software code has been designed and implemented. The software comes with maps and is very easy to be used because assessment scenarios can be done through diagrams. Software verification was performed by comparing its result to SRS-19 and CROM software calculation results. Dose estimated by SRS-19 are higher compared to the result of developed software. However, these are still acceptable since dose estimation in SRS-19 is based on conservative approach. On the other hand, compared to CROM software, the same results for three scenarios and a non-significant difference of 2.25% in another scenario were obtained. These results indicate the correctness of our implementation and implies that the developed software is ready for use in real scenario. In the future, the addition of various features and development of new model need to be done to improve the capability of software that has been developed. Keywords: Radiation dose assessment, software code, radioactive discharge, environment, IAEA SRS-19. PERANCANGAN, IMPLEMENTASI DAN VERIFIKASI PERANGKAT LUNAK UNTUK KAJIAN DOSIS RADIASI BERBASIS MODEL LINGKUNGAN GENERIK SEDERHANA. Kajian dosis radiasi untuk mengetahui potensi dampak radiologi akibat pengoperasian berbagai instalasi yang ada di dalam kawasan instalasi nuklir sangat diperlukan untuk menjamin keselamatan masyarakat dan lingkungan. Metode sederhana berbasis model lingkungan generik untuk menghitung dosis radiasi yang disebabkan oleh pelepasan zat radioaktif ke lingkungan telah diterbitkan oleh International Atomic Energy Agency (IAEA) sebagai Safety Report Series No. 19 (SRS-19). Untuk lebih memudahkan penerapan metode tersebut, maupun sebagai dasar kajian serta pengembangan metode yang lebih kompleks, maka dilakukan perancangan dan implementasi perangkat lunak terkait berbasis open-source. Perangkat lunak yang dibuat telah dilengkapi dengan peta serta sangat mudah digunakan karena skenario kajian dosis radiasi dapat dibuat melalui diagram. Verifikasi melalui komparasi dengan nilai hasil perhitungan di SRS-19 serta hasil perhitungan perangkat lunak CROM. Hasil komparasi dengan SRS-19 menunjukkan adanya perbedaan pada empat skenario yang diuji. Hal ini wajar karena pada SRS-19 perhitungan dilakukan dengan pendekatan konservatif sehingga nilai dosis yang dihasilkan lebih besar jika dibandingkan dengan perangkat lunak yang dikembangkan. Selanjutnya, pada komparasi dengan perangkat lunak CROM, keduanya menunjukkan hasil yang sama pada tiga skenario serta perbedaan yang tidak siginifikan yaitu sebesar 2,25% pada satu skenario. Hasil tersebut menunjukkan bahwa implementasi perangkat lunak telah berhasil dan perangkat lunak tersebut dapat digunakan untuk kajian yang nyata. Kedepannya, penambahan berbagai fitur serta pengembangan model kajian perlu dilakukan untuk meningkatkan kemampuan perangkat lunak yang telah dibuat. Kata kunci: Kajian dosis radiasi, perangkat lunak, lepasan zat radioaktif, lingkungan, IAEA SRS-19

A postulated loss of coolant accident (LOCA) shall be analyzed to assure the safety of a research reactor. The analysis of such accident could be performed using best estimate thermal-hydraulic codes, such as RELAP5. This study focuses on analysis of LOCA in TRIGA-2000 due to pipe and beam tube break. The objective is to understand the effect of break size and the actuating time of the emergency core cooling system (ECCS) on the accident consequences and to assess the safety of the reactor. The analysis is performed using RELAP/SCDAPSIM codes. Three different break size and actuating time were studied. The results confirmed that the larger break size, the faster coolant blow down. But, the siphon break holes could prevent the core from risk of dry out due to siphoning effect in case of pipe break. In case of beam tube rupture, the ECCS is able to delay the fuel temperature increased where the late actuation of the ECCS could delay longer. It could be concluded that the safety of the reactor is kept during LOCA throughout the duration time studied. However, to assure the integrity of the fuel for the long term, the cooling system after ECCS last should be considered. Keywords: safety analysis, LOCA, TRIGA, RELAP5 STUDI PARAMETRIK LOCA DI TRIGA-2000 MENGGUNAKAN RELAP5/SCDAP. Kecelakaan kehilangan air pendingin (LOCA) harus dianalisis untuk menjamin keselamatan suatu reaktor riset. Analisis LOCA dapat dilakukan menggunakan perhitungan best-estimate seperti RELAP5. Penelitian ini menekankan pada analisis LOCA di TRIGA-2000 akibat pecahnya pipa dan tabung berkas. Tujuan penelitian adalah memahami efek ukuran kebocoran dan waktu aktuasi sistem pendingin teras darurat (ECCS) pada sekuensi kejadian dan mengkaji keselamatan reaktor. Analisis dilakukan menggunakan program perhitungan RELAP/SCDAPSIM. Tiga ukuran kebocoran dan waktu aktuasi ECCS berbeda dipilih sebagai parameter dalam studi ini. Hasil perhitungan mengonfirmasi bahwa semakin besar ukuran kebocoran, semakin cepat pengosongan tangki reaktor. Lubang siphon breaker dapat mencegah air terkuras dalam hal kebocoran pada pipa. Sedang dalam hal kebocoran pada beam tube, ECCS mampu memperlambat kenaikan temperatur bahan bakar. Dari studi ini dapat disimpulkan bahwa keselamatan reaktor dapat terjaga pada kejadian LOCA, namun pendinginan jangka panjang perlu dipertimbangkan untuk menjaga integritas bahan bakar.Kata kunci: analisis keselamatan, LOCA, TRIGA, RELAP5

HTGR is one of the next generation reactor types. HTGR is currently considered as one of the leading reactors for the future nuclear power plant. The steam generator is one of the main components in HTGR as well as in RDE. In the steam generator, the heat is transferred by high temperature helium gas in the shell side to water in the tube side to generate the superheated steam. the purpose of this work is to design the operating parameter of outlet temperature and void fraction of steam based on feed water mass flow rate and inlet temperature variations in RDE steam generator. In this work, the Chemcad program was used. Both inlet and outlet temperature of helium gas have been set up as boundary conditions. The result shows that using the mass flow rate of 4.3 kg/s - 4.8 kg/s and water inlet temperature of 110 o C - 160 o C, the superheated steam outlet temperature (void fraction = 1.0) is obtained in the range of 275.5 o C – 600 o C.This analysis is beneficial to assess 10 MW RDE design especially in the steam generator system operating parameters. Keywords: outlet temperature, void fraction, superheated steam, RDE steam generator ANALISIS DESAIN PARAMETER OPERASI UNTUK TEMPERATUR KELUARAN DAN FRAKSI UAP PADA PEMBANGKIT UAP RDE. Reaktor daya HTGR adalah salah satu tipe reaktor generasi lanjut. HTGR saat ini merupakan desain reaktor yang dipertimbangkan untuk pembangkit listrik unggulan dimasa mendatang. Pembangkit uap merupakan salah satu komponen utama pada HTGR begitu pula pada RDE. Di dalam pembangkit uap, panas dari gas helium temperatur tinggi pada sisi shell di transfer ke air pada sisi tube pembangkit uap untuk menghasilkan uap lewat jenuh. Tujuan analisis ini adalah mendesain parameter operasi terhadap temperatur keluaran dan fraksi uap berdasarkan variasi laju alir massa air umpan dan temperatur masuk pada RDE. Dalam analisis digunakan program Chemcad, temperatur gas helium masuk dan keluar ditentukan sebagai kondisi batas. Hasil menunjukkan bahwa dengan menggunakan laju alir massa 4,3 kg/detik - 4,8 kg/detik dan temperatur masukan air umpan dari 110 o C -160 o C dapat diperoleh uap lewat jenuh (fraksi uap= 1,0) pada temperatur keluaran dalam rentang 275,5 o C - 600 o C. Analisis ini berguna untuk memberikan kajian desain RDE 10 MW khususnya parameter operasi sistem pembangkit uap. Kata-kata kunci: temperatur keluaran, fraksi uap, uap lewat jenuh, pembangkit uap RDE

LEMBAR ABSTRAK + INDEKS

Missing values are problems in data evaluation. Missing values analysis can resolve the problem of incomplete data that is not stored properly. The missing data can reduce the precision of calculation, since the amount of information is incomplete. The purpose of this study is to implement missing values handling method for systems/components maintenance historical data evaluation in RSG GAS. Statistical methods, such as listwise deletion and mean substitution, and machine learning (KNNI), were used to determine the missing data that correspond to the systems/components maintenance historical data. Mean substitution and KNNI methods were chosen since those methods do not require the formation of predictive models for each item which is experiencing missing data. Implementation of missing data analysis on systems/components maintenance data using KNNI method results in the smallest RMSE value. The result shows that KNNI method is the best method to handle missing value compared with listwise deletion or mean substitution. Keywords: missing value, data evaluation, alghorithm, implementation IMPLEMENTASI METODE PENANGANAN DATA HILANG UNTUK MENGEVALUASI DATA SEJARAH PERAWATAN SISTEM/ KOMPONEN . Data hilang merupakan masalah dalam melakukan evaluasi data. Analisis data hilang dapat menyelesaikan permasalahan ketidaklengkapan data yang tidak tersimpan dengan baik . Data yang hilang akan memperkecil presisi dari perhitungan, dikar enakan jumlah informasi yang tidak lengkap. Tujuan dari penelitian ini adalah implementasi metode penanganan data hilang untuk evaluasi data sejarah perawatan sistem/ komponen RSG GAS. Metodologi yang digunakan untuk menentukan data hilang yang berhubungan dengan data sejarah perawatan sistem/komponen adalah s tatistics , listwise deletion dan mean substitution , dan machine learning (KNNI). Metode mean substitution dan KNNI dipilih karena metode ini tidak memerlukan informasi untuk pembentukan model prediksi untuk setiap item y a ng mengandung data hilang . I mplementasi analisis data hilang pada data perawatan sistem/komponen menggunakan metode KNNI menghasilkan nilai RMSE terkecil. Hasil ini menunjukan bahwa metode KNNI merupakan metode terbaik untuk menangani data hilang dibanding dengan listwise deletion atau mean substitution . Kata kunci: data hilang, evaluasi data, algoritma, implementasi

Cover + TOC + Preface

In the in-service inspection conducted to G.A. Siwabessy reactor’s primary cooling system pipe, it was found the presence of inhomogenity inside of welding part. To verify whether the inhomogenity could be tolerated or not, comparative data from welding pre-service inspection is needed. Unfortunately, this weld wasn’t covered in pre-service inspection. Therefore, this inhomogenity needs to be analyzed. The purpose of this study is to evaluate the stress intensity factor of the inhomogenity, whether it is within a limit value or not and to predict the crack growth. Analysis were performed based on fracture mechanics theory using parameter of stress intensity factor. Two models were used for calculation approach that are plane crack model and semi-elliptic crack model. Hence, in order to predict the length of inhomogenity in the future, crack growth calculations were performed. The results showed that stress intensity values from both two models are remain below fracture toughness value of pipe’s material. Besides that, stress intensity factor from plane crack model is higher than those from semi-elliptic crack model. Under consideration that inhomogenity has an arc shape in actual, thus, stress intensity factor from this inhomogenity still low enough compare to the fracture toughness. Crack growth calculation’s results showed that after 300 th cycle of loading, the length of inhomogenity reaches approximately 2 mm. Based on operation data of G.A. Siwabessy reactor, 300 cycle number is corresponds to 30 years operation. Based on these results it could be concluded that the presence of inhomogenity in the welding part does not affect the structure’s integrity of piping system. Keywords : Inhomogenity, fracture mechanics, fracture toughness, stress intensity factor, crack growth ABSTRAK Pada pelaksanaan in-service inspection terhadap perpipaan sistem pendingin primer reaktor G.A. Siwabessy diketahui adanya inhomogenitas pada salah satu sambungan lasan pipa. Untuk memverifikasi apakah inhomogenitas ini dapat ditoleransi atau tidak, diperlukan data pembanding hasil pemeriksaan lasan pada saat fabrikasi. Namun, ternyata pada saat fabrikasi, sambungan lasan ini tidak mengalami pemeriksaan. Oleh karena itu, dalam rangka menetapkan apakah keberadaan inhomogentitas ini dapat ditoleransi atau tidak perlu dilakukan analisis terhadap inhomogenitas tersebut. Tujuan penelitian ini adalah untuk melakukan evaluasi stress intensity factor inhomogenitas di dalam pipa apakah masih berada di dalam batas nilai dan untuk memprediksi perambatan retak. Analisis dilakukan berdasarkan teori fracture mechanics dengan menghitung stress intensity factor inhomogenitas. Dalam perhitungan ini digunakan dua model untuk pendekatan, yaitu model retak planar dan model retak semi-ellips. Selanjutnya, untuk memprediksi panjang inhomogenitas di masa yang akan datang, dilakukan juga simulasi perambatan retak. Hasil-hasil analisis memperlihatkan bahwa nilai stress intensity factor berdasarkan model retak bentuk planar dan retak bentuk semi ellips masih jauh di bawah nilai fracture toughness material pipa. Selain itu, nilai yang dihasilkan berdasarkan model retak bentuk planar lebih besar dibandingkan dengan model retak bentuk semi ellips. Mengingat bentuk inhomogenitas yang berupa busur lingkaran, maka nilai stress intensity factor yang sesungguhnya dari inhomogenitas tersebut jauh lebih kecil dibandingkan dengan nilai fracture toughness. Sementara itu, untuk hasil simulasi perambatan retak menunjukkan bahwa pada siklus pembebanan ke-300 memberikan panjang sekitar 2 mm. Berdasarkan data operasi reaktor G.A. Siwabessy, jumlah siklus sebanyak 300 kali setara dengan pengoperasian reaktor selama 30 tahun. Berdasarkan dua hasil tersebut dapat disimpulkan bahwa keberadaan inhomogenitas pada sambungan lasan tidak berpengaruh terhadap integritas struktur sistem perpipaan. Kata kunci : Inhomogenitas, fracture mechanincs, fracture toughness, stress intensity factor, pertumbuhan retak

Operator roles and intervene actions on the operation of gas cooled reactor would be different compared to their roles in other reactor types. Analysis of operator performance and the influencing factors can be conducted comprehensively in Human Reliability Analysis (HRA). Using HRA, the impact of human errors on the system and the ways to reduce human error impact and frequency can be idenfified. The paper discusses factors influencing reactor operator performance to response to the cooling accident of the high temperature gas cooled reactor (HTGR). Analysis and qualification of influencing factors, which are performance shaping factors (PSF), were conducted based on time reliability curve and Cognitive Reliability and Error Analysis Method (CREAM). Based on time reliability curve, results showed that time variable contributes to the improvement of operator performance (PSF

EVALUATION ON MECHANICAL FRACTURE OF PWR PRESSURE VESSEL AND MODELING BASED ON NEURAL NETWORK. The important component of the PWR is a pressure vessel. The material resistance in the pressure vessel needs to be evaluated. One way of evaluation is by the mechanical fracture analysis. The modeling needs to know the phenomena of the analysis result in general. A number of researches have been completed on the calculation of mechanical fracture in the pressure vessel with an internal load. The mechanical fracture was modeled using a neural network approach. In relation to the material resistance of the pressure vessel, which is used in PWR AP1000, the material must be evaluated because of the effect of the load. The modeling is needed to predict the effect of the load. The aim of this study is to evaluate the material resistance through mechanical fracture analysis because of the influence load on the pressure vessel on PWR AP1000. The material, which was observed, is SA 508. This analysis consists of the calculation of stress intensity factor and J-integral with some load at the crack propagation position. The fracture mechanic was analyzed by finite element simulation. The result of Stress Intensity factor and J-Integral was compared with fracture toughness to know the durability of the material. The modeling of J-Integral and Stress Intensity Factor were obtained for some load based on neural network approach. Keywords: Material resistance, mechanical fracture, neural network, PWR, pressure vessel, crack propagation. ABSTRAK EVALUASI FRAKTUR MEKANIK PADA BEJANA TEKAN PWR DAN PEMODELAN BERBASIS NEURAL NETWORK. Komponen penting dari PWR adalah bejana tekan. Ketahanan bahan di bejana tekan perlu dievaluasi. Salah satu cara adalah dengan analisis fraktur mekanik. Pemodelan diperlukan untuk mengetahui fenomena hasil analisis pada umumnya. Terdapat penelitian untuk perhitungan fraktur mekanik dalam bejana tekan dengan beban internal. Penelitian lain adalah hasil dari fraktur mekanik dimodelkan menggunakan pendekatan jaringan syaraf. Sehubungan dengan ketahanan material dari bejana tekan yang digunakan dalam PWR AP1000, bahan harus dievaluasi karena efek dari beban. Pemodelan diperlukan untuk memprediksi pengaruh beban pada bahan dalam bejana tekan. Tujuan dari penelitian ini adalah untuk mengevaluasi ketahanan material melalui analisis fraktur mekanik karena pengaruh beban pada bejana tekan. Bahan yang diamati, adalah SA 508. Analisis ini terdiri dari perhitungan faktor intensitas tegangan dan J-integral dengan beberapa beban pada posisi perambatan retak. Fraktur mekanik dianalisis dengan metode elemen hingga. Hasil faktor intensitas tegangan dan J-Integral dibandingkan dengan ketangguhan patah untuk mengetahui daya tahan material. Pemodelan J-Integral dan faktor intensitas stres diperoleh untuk beberapa beban berdasarkan jaringan saraf. Kata kunci: Ketahanan bahan, teknik patahan, jaringan syaraf, PWR, bejana tekan, perambatan retak.