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Typical TRIGA neutronic parameters calculation has been performed using SRAC 2006 with 3 dimensions modeling. Calculation of the criticality and power distribution with the SRAC2006 computer program is carried out in 2 stages. The first step is carried out by calculating the criticality of each fuel element and control rod using the PIJ module. Then, the second is the 3-dimensional core criticality calculation with the CITATION module. There are three core configuration scenarios calculated in this study. Calculated neutronic parameters are multiplication factor (K-Eff), radial and axial power peaking factor (PPF), radial and axial neutron distribution, and reactor core power density. The objective of this study is to conduct a safety analysis of the TRIGA reactor core configuration scenarios. The calculation results of the multiplication factor for reactor core configuration scenarios 1, 2, and 3 are 1.02996, 1.02761, and 1.02776. And the radial power peaking factors are 1.64, 1.68, and 1.89. Whereas, for axial Power Peaking Factor, it is still within the operating limits and conditions value range of 1.2. For the radial PPF, it can be concluded that the core configuration scenario 1 and scenario 2 can be considered as the safe core configuration because their radial PPF values are close to the operating limit and condition for this typical TRIGA reactor. The reactor core configuration scenario 3 has radial Power Peaking Factor values that exceeds operating limits and conditions for the TRIGA reactor so this scenario cannot be used to operate the reactor. The closer the radial PPF to the operating limit and condition value, the safer the reactor core configuration.

Center for Applied Nuclear Science and Technology - The National Nuclear Energy Agency plans to modify the TRIGA 2000 Bandung Reactor to Bandung Conversion Reactor by replacing standard TRIGA cylindrical fuel with domestic- made plate type fuel. The consequence of this modification plan is that we must modify the transfer cask. An adequate transfer cask is needed so that the radiation exposure emitted from transfer cask is enough low so that the workers, who handling the spent fuel transfer can work safely. The transfer cask was constructed from two concentric cylindrical steel shells with a bolted top cover plate and a welded bottom. The annulus formed by two shells is filled with lead cast, which functions as a gamma shield.. The results of calculations using the ORIGEN2 software for 18 energy groups, 26% fuel element and 1 day decay time, were used as input for the QADS-SCALE-4.3 software. The transfer cask design was chosen with a thickness of 27 cm, an inner diameter of 14 cm, an outer diameter of 68 cm and a height of 140 cm made of lead. The design provided satisfactory shielding capabilities within the limit of the surface dose rates.Center for Applied Nuclear Science and Technology - The National Nuclear Energy Agency plans to modify the TRIGA 2000 Bandung Reactor to Bandung Conversion Reactor by replacing standard TRIGA cylindrical fuel with domestic- made plate type fuel. The consequence of this modification plan is that we must modify the transfer cask. An adequate transfer cask is needed so that the radiation exposure emitted from transfer cask is enough low so that the workers, who handling the spent fuel transfer can work safely. The transfer cask was constructed from two concentric cylindrical steel shells with a bolted top cover plate and a welded bottom. The annulus formed by two shells is filled with lead cast, which functions as a gamma shield.. The results of calculations using the ORIGEN2 software for 18 energy groups, 26% fuel element and 1 day decay time, were used as input for the QADS-SCALE-4.3 software. The transfer cask design was chosen with a thickness of 27 cm, an inner diameter of 14 cm, an outer diameter of 68...

The COOLOD-N2 and RELAP5 codes have been applied for a steady-state thermal-hydraulic safety analysis of the 2 MW TRIGA Plate core design at the BATAN, Indonesia. In order to get the best prediction, the maximum allowable axial node division was applied in COOLOD-N2 input model and the RELAP5 input nodalization was adjusted to match the COOLOD-N2 input model. The main objective of this research is to confirm the safety margins of different safety related parameter by steady-state calculation at full power level (2 MW). Therefore, we have calculated the coolant temperature, cladding surface temperature, fuel center temperature, departure from nucleate boiling ratio, and onset of nucleate boiling temperature profile across the hottest channel. The most significant conclusion is that all obtaining values indicate the reactor has enough safety margin.

Bandung TRIGA Plate reactor core was designed to have 5 x 5 size by using 16 fuel elements, 4 control elements and 5 irradiation facilities. The irradiation facility is located at A1, A5, C3, E1, and E5. This study was conducted in order to determine the effect of the center irradiation position at C3 position to the fuel element temperature distribution at the reactor core. Simulation study was conducted using Fluent computer program. The irradiation facility greatly influences coolant flow rate and the distribution of coolant flow into the reactor core, which greatly affects to the reactor core temperature. The results of the study showed a significant increase in surface temperature of the fuel element when center irradiation position at C3 position was discharged. Therefore, it is recommended if at C3 position there is no material to be irradiated, then the position shall be filled using a solid material. The purpose of this closure is to reduce the cooling flow rate passing through the void at C3 position. However, neutronic design analysis should also consider the type of a material that will be used to fill the void, so that it does not affect to the neutron flux in the reactor core.

Reactor TRIGA PUSPATI (RTP) Mark II type undergoes safe operation for more than 30 years and the only research reactor exists in Malaysia. The main safety feature of Instrumentation and Control (I&C) system design is such that any failure in the electronic, or its associated components, does not lead to an uncontrolled rate of reactivity. The existed controller using feedback approach to control the reactor power. This paper introduces proposed controllers such as Model Reference Adaptive Control (MRAC) and Proportional Integral Derivatives (PID) controller for the RTP simulation. In RTP, the most important considered parameter is the reactor power and act as nervous system. To design a controller for complex plant like RTP is quite difficult due to high cost and safety factors cause by the failure of the controller. Furthermore, to overcome these problems, a simulator can be used to replace functions the hardware and test could then be simulated using this simulator. In order to find the best controller, several controllers were proposed and the result will be analysed for study the performances of the controller. The output result will be used to find out the best RTP power controller using MATLAB/Simulink and gives result as close as the real RTP performances. Currently, the structures of RTP was design using MATLAB/Simulink tool that consist of fission chamber, controller, control rod position, height-to-worth of control rods and a RTP model. The controller will control the control rod position to make sure that the reactivity still under the limitation parameter. The results given from each controller will be analysed and validated through experiment data collected from RTP.

TRIGA Bandung is a research nuclear reactor owned by Indonesia, located in Bandung with a power of 2 MWth. Nuclear research reactor TRIGA Bandung is used as a center for applied research and development in the field of application of the nuclear technologies. TRIGA Bandung is currently still using a cylindrical fuel element, this raises a new problem - the limited number of existing fuel element. The purpose of this research is the development of the preliminary core design of a nuclear research reactor TRIGA Bandung using fuel element plate MTR. The research method is modeling and simulation the preliminary design core of nuclear research reactor TRIGA Bandung using comparative method of porous media and non-porous media with CFD code. This research shows the velocity flow and temperature distribution and the influence of pressure from the comparison method of k-e standard model and porous media model at the preliminary design in the core area of TRIGA Bandung research reactor with fuel element plate MTR.

An electrical power systems are exposed to different types of power quality disturbances. Investigation and monitoring of power quality are necessary to maintain accurate operation of sensitive equipment especially for nuclear installations. This paper will discuss the power quality problems observed at the electrical sources of PUSPATI TRIGA Reactor (RTP). Assessment of power quality requires the identification of any anomalous behavior on a power system, which adversely affects the normal operation of electrical or electronic equipment. A power quality assessment involves gathering data resources; analyzing the data (with reference to power quality standards) then, if problems exist, recommendation of mitigation techniques must be considered. Field power quality data is collected by power quality recorder and analyzed with reference to power quality standards. Normally the electrical power is supplied to the RTP via two sources in order to keep a good reliability where each of them is designed to carry ...

We study the amount of temperature feedback on reactivity for the three types of TRIGA fuel i.. ST8, ST12 and LEU fuel, are used in the TRIGA MARK II reactor in Malaysia Nuclear Agency. We employ WIMSD-5B for the calculation of kin f for a single TRIGA fuel surrounded by water. Typical calculations of TRIGA fuel reactivity are usually limited to ST8 fuel, but in this paper our investigation extends to ST12 and LEU fuel. We look at the kin f of our model at various fuel temperatures and calculate the amount reactivity removed. In one instance, the water temperature is kept at room temperature of 300K to simulate sudden reactivity increase from startup. In another instance, we simulate the sudden temperature increase during normal operation where the water temperature is approximately 320K while observing the kin f at various fuel temperatures. For accidents, two cases are simulated. The first case is for water temperature at 370K and the other is without any water. We observe that the higher Uranium content fuel such as the ST12 and LEU have much smaller contribution to the reactivity in comparison to the often studied ST8 fuel. In fact the negative reactivity coefficient for LEU fuel at high temperature in water is only slightly larger to the negative reactivity coefficient for ST8 fuel in void. The performance of ST8 fuel in terms of negative reactivity coefficient is cut almost by half when it is in void. These results are essential in the safety evaluation of the reactor and should be carefully considered when choices of fuel for core reconfiguration are made.

Since its first criticality in 1982, PUSPATI TRIGA Reactor (RTP) has been operated for more than 30 years. As RTP become older, ageing problems have been seen to be the prominent issues. In addressing the ageing issues, an Ageing Management (AgeM) database for managing related ageing matters was systematically developed. This paper presents the development of AgeM database taking into account all RTP major Systems, Structures and Components (SSCs) and ageing mechanism of these SSCs through the system surveillance program.

A high performance neutron diffractometer called Savar Neutron Diffractometer (SAND) was built and installed at radial beam port-2 of TRIGA Mark II research reactor at AERE, Savar, Dhaka, Bangladesh. Structural studies of materials are being done by this technique to characterize materials crystallograpohically and magnetically. The micro-structural information obtainable by neutron scattering method is very essential for determining its technological applications. This technique is unique for understanding the magnetic behavior in magnetic materials. Ceramic, steel, electronic and electric industries can be benefited from this facility for improving their products and fabrication process. This instrument consists of a Popovicimonochromator with a large linear position sensitive detector array. The monochromator consists of nine blades of perfect single crystal of silicon with 6mm thickness each. The monochromator design was optimized to provide maximum flux on 3mm diameter cylindrical sample with a relat...

A simple continuous flow (SCF) cryostat was designed to investigate the neutron moderation of alumina in high temperature co-ceramic (HTCC) and polymeric materials such as Teflon under TRIGA neutron environment using a reflected neutron beam from a monochromator. Cooling of the cryostat will be carried out using liquid nitrogen. The cryostat will be built with an aluminum holder for moderator within stainless steel cylinder pipe. A copper thermocouple will be used as the temperature sensor to monitor the moderator temperature inside the cryostat holder. Initial measurements of neutron spectrum after neutron passing through the moderating materials have been carried out using a neutron spectrometer.

The 1-MW PUSPATI Research Reactor (RTP) is the one and only nuclear pool type research reactor developed by General Atomic (GA) in Malaysia. It was installed at Malaysian Nuclear Agency and has reached the first criticality on 8 June 1982. Based on the initial core which comprised of 80 standard TRIGA fuel elements, the very fundamental thermal hydraulic model was investigated during steady state operation using the PARET-code. The main objective of this paper is to determine the variation of temperature profiles and Departure of Nucleate Boiling Ratio (DNBR) of RTP at full power operation. The second objective is to confirm that the values obtained from PARET-code are in agreement with Safety Analysis Report (SAR) for RTP. The code was employed for the hot and average channels in the core in order to calculate of fuel’s center and surface, cladding, coolant temperatures as well as DNBR’s values. In this study, it was found that the results obtained from the PARET-code showed that the thermal hydraulic parameters related to safety for initial core which was cooled by natural convection was in agreement with the designed values and safety limit in SAR.

Some of the oldest TRIGA fuel in the Malaysian Reaktor TRIGA PUSPATI (RTP) is approaching the limit of its end of life with burn-up of around 20%. Hence it is prudent for us to start planning on the replacement of the fuel in the reactor and other derivative activities associated with it. In this regard, we need to understand all of the risk associated with such operation and one of them is to predict the radioactivity of the fuel, so as to estimate the safety of our working conditions. The radioactivity of several fuels are measured and compared with simulation results to confirm the burnup levels of the selected fuels. The radioactivity measurement are conducted inside the water tank to reduce the risk of exposure and in this case the detector wrapped in plastics are lowered under water. In nuclear power plant, the general practice was to continuously burn the fuel. In research reactor, most operations are based on the immediate needs of the reactor and our RTP for example operate periodically. By integrating the burnup contribution for each core configuration, we simplify the simulation of burn up for each core configuration. Our results for two (2) fuel however indicates that the dose from simulation underestimate the actual dose from our measurements. Several postulates are investigated but the underlying reason remain inconclusive.

Nuclear Track Methodology (NTM), based on the neutron-proton interaction is one often employed alternative for neutron detection. In this paper we apply NTM to determine the Triga Mark III reactor operating power and neutron flux. The facility nuclear core, loaded with 85 Highly Enriched Uranium as fuel with control rods in a demineralized water pool, provide a neutron flux around 2 × 1012 n cm−2 s−1, at the irradiation channel TO-2. The neutron field is measured at this channel, using Landauer® PADC as neutron detection material, covered by 3 mm Plexiglas® as converter. After exposure, plastic detectors were chemically etched to make observable the formed latent tracks induced by proton recoils. The track density was determined by a custom made Digital Image Analysis System. The resulting average nuclear track density shows a direct proportionality response for reactor power in the range 0.1-7 kW. We indicate several advantages of the technique including the possibility to calibrate the neutron flux dens...

Spent fuel transfer cask is used to transfer a spent fuel from the reactor tank to the spent fuel storage or for spent fuel inspection. Typically, the cask made from steel cylinders that are either welded or bolted closed. The cylinder is enclosed with additional steel, concrete, or other material to provide radiation shielding and containment of the spent fuel. This paper will discuss the Conceptual Design of fuel transfer cask for Reactor TRIGA Puspati (RTP).

ORIGEN 2.2 are employed to obtain data regarding γ source term and the radio-activity of irradiated TRIGA fuel. The fuel composition are specified in grams for use as input data. Three types of fuel are irradiated in the reactor, each differs from the other in terms of the amount of Uranium compared to the total weight. Each fuel are irradiated for 365 days with 50 days time step. We obtain results on the total radioactivity of the fuel, the composition of activated materials, composition of fission products and the photon spectrum of the burned fuel. We investigate the differences of results using BWR and PWR library for ORIGEN. Finally, we compare the composition of major nuclides after 1 year irradiation of both ORIGEN library with results from WIMS. We found only minor disagreements between the yields of PWR and BWR libraries. In comparison with WIMS, the errors are a little bit more pronounced. To overcome this errors, the irradiation power used in ORIGEN could be increased a little, so that the differences in the yield of ORIGEN and WIMS could be reduced. A more permanent solution is to use a different code altogether to simulate burnup such as DRAGON and ORIGEN-S. The result of this study are essential for the design of radiation shielding from the fuel.

In this paper, spent ion exchange resin generated from PUSPATI TRIGA reactor (RTP) in Malaysian Nuclear Agency were characterized based on the water content, radionuclide content and radionuclide leachability. The result revealed that the water content in the spent resin is 48%. Gamma spectrometry analysis indicated the presence of 134Cs, 137Cs, 152Eu, 54Mn, 58Co, 60Co and 65Zn. The leachability test shows a small concentrations (

The purpose of this preliminary work is to study for thickness changes in objects using neutron radiography. In doing the project, the technique for the radiography was studied. The experiment was done at NUR-2 facility at TRIGA research reactor in Malaysian Nuclear Agency, Malaysia. Test samples of varying materials were used in this project. The samples were radiographed using direct technique. Radiographic images were recorded using Nitrocellulose film. The films obtained were digitized to processed and analyzed. Digital processing is done on the images using software Isee!. The images were processed to produce better image for analysis. The thickness changes in the image were measured to be compared with real thickness of the objects. From the data collected, percentages difference between measured and real thickness are below than 2%. This is considerably very low variation from original values. Therefore, verifying the neutron radiography technique used in this project.

Malaysian nuclear research reactor, the PUSPATI TRIGA Reactor, reached its first criticality in 1982, and since then, it has been serving for more than 30 years for training, radioisotope production and research purposes. Realizing the age and the need for its decommissioning sometime in the future, a ground basis of assessment and an elaborative project management need to be established, covering the entire process from termination of reactor operation to the establishment of final status, documented as the Decommissioning Plan. At international level, IAEA recognizes the absence of Decommissioning Plan as one of the factors hampering progress in decommissioning of nuclear facilities in the world. Throughout the years, IAEA has taken initiatives and drawn out projects in promoting progress in decommissioning programmes, like CIDER, DACCORD and R2D2P, for which Malaysia is participating in these projects. This paper highlights the concept of Decommissioning plan and its significances to the Agency. It wil...

After operating for more than 25 years, some of the Reaktor TRIGA Puspati (RTP) fuel elements would have been depleted. A few addition and fuel reconfiguration exercises have to be conducted in order to maintain RTP capacity. Presently, RTP spent fuels are stored at the storage area inside RTP tank. The need to transfer the fuel element outside of RTP tank may be prevalence in the near future. The preparation shall be started from now. A fuel element transfer cask has been designed according to the recommendation by the fuel manufacturer and experience of other countries. A modelling using MCNP code has been conducted to analyse the design. The result shows that the design of transfer cask fuel element is safe for handling outside the RTP tank according to recent regulatory requirement.

To accomplish safety requirements, a set of actions has to be performed following the recommendations of the IAEA safety series 35 applied to research reactor. Such actions are considered in modernization of the old system, improving the core cooling system and safety evaluations. Due to the complexity of the process and the difficulty in putting the apparatus in the reactor core, analytical and experimental study on the determination of flow and temperature distribution in the whole coolant channel are difficult to be done. In the present work, a numerical study of flow and temperature distribution in the coolant channel of TRIGA 2000 has been carried out using CFD package. For this study, simulations were carried out on 3‐D tested model. The model consists of the reactor tank, thermal and thermalizing column, reflector, rotary specimen rack, chimney, fuel element, primary pipe, diffuser, beam tube and a part of the core are constructed by 1.50 million unstructured tetrahedral cell elements. The results show that for the initial condition (116 fuel elements in the core) and for the inlet temperature of 24°C and the primary velocity of 5.6 m/s, there no boiling phenomena occur in the coolant channel. Due to this result, it is now possible to improve the core cooling system of TRIGA 2000 reactor. Meanwhile, forced flow from the diffuser system only affected the flow pattern in the outside of chimney and put on a small effect to the fluid flow’s velocity in the inside of chimney.

In order to study neutron‐rich nuclides far from the valley of stability as well as long‐lived actinoids the double Penning‐trap mass spectrometer TRIGA‐TRAP has been recently installed at the research reactor TRIGA Mainz. Short‐lived neutron‐rich fission products are produced by thermal neutron‐induced fission of an actinoid target installed close to the reactor core. A helium gas‐jet system with carbon aerosol particles is used to extract the fission products to the experiment. The Penning trap system has already been commissioned. Off‐line mass measurements are routinely performed using a recently developed laser ablation ion source, and the gas‐jet system has been tested. An overview of the experiment and current status will be given.

Common energy crisis has modified the national energy policy which is in the beginning based on natural resources becoming based on technology, therefore the capability to understanding the basic and applied science is needed to supporting those policies. National energy policy which aims at new energy exploitation, such as nuclear energy is including many efforts to increase the safety reactor core condition and optimize the related aspects and the ability to build new research reactor with properly design. The previous analysis of the modification TRIGA 2000 Reactor design indicates that forced convection of the primary coolant system put on an effect to the flow characteristic in the reactor core, but relatively insignificant effect to the flow velocity in the reactor core. In this analysis, the lid of reactor core is closed. However the forced convection effect is still presented. This analysis shows the fluid flow velocity vector in the model area without exception. Result of this analysis indicates that in the original design of TRIGA 2000 reactor, there is still forced convection effects occur but less than in the modified TRIGA 2000 design.

Malaysian Nuclear Agency’s (Nuclear Malaysia) Small Angle Neutron Scattering (SANS) facility—(MYSANS)—is utilizing low flux of thermal neutron at the agency’s 1 MW TRIGA reactor. As the design nature of the 8 m SANS facility can allow object resolution in the range between 5 and 80 nm to be obtained. It can be used to study alloys, ceramics and polymers in certain area of problems that relate to samples containing strong scatterers or contrast. The current SANS system at Malaysian Nuclear Agency is only capable to measure Q in limited range with a PSD (128×128) fixed at 4 m from the sample. The existing reactor hall that incorporate this MYSANS facility has a layout that prohibits the rebuilding of MYSANS therefore the position between the wavelength selector (HOPG) and sample and the PSD cannot be increased for wider Q range. The flux of the neutron at current sample holder is very low which around 103 n/cm2/sec. Thus it is important to rebuild the MYSANS to maximize the utilization of neutron. Over the ...

In this paper I shall give a short history of the development of neutron scattering at the Research Centre for Nuclear Techniques (PPTN), in Bandung, and the early development of a more advanced facilities at the Neutron Scattering Laboratory (NSL BATAN), Centre of Technology for Nuclear Industrial Materials, in Serpong. The first research reactor in Indonesia was the TRIGA MARK II in Bandung, which became operational in 1965, with a power of 250 KW, upgraded to 1 MW in 1971, and to 2 MW in 2000. The neutron scattering activities was started in 1967, with the design and construction of the first powder diffractometer, and put in operation in 1970. It was followed by the second instrument, the filter detector spectrometer built in 1975 in collaboration with the Bhabha Atomic Research Centre (BARC), India. A powder diffractometer for magnetic studies was built in 1980, and finally, a modification of the filter detector spectrometer to measure textures was made in 1986. A brief description of the design and ...

The SUSEE space reactor system uses existing nuclear fuels and the standard steam cycle to generate electrical and thermal power for a wide range of in‐space and surface applications, including manned bases, sub‐surface mobile probes to explore thick ice deposits on Mars and the Jovian moons, and mobile rovers. SUSEE cycle efficiency, thermal to electric, ranges from ∼20 to 24%, depending on operating parameters. Rejection of waste heat is by a lightweight condensing radiator that can be launched as a compact rolled‐up package and deployed into flat panels when appropriate. The 50 centimeter diameter SUSEE reactor can provide power over the range of 10 kW(e) to 1 MW(e) for a period of 10 years. Higher power outputs are possible using slightly larger reactors. System specific weight (reactor, turbine, generator, piping, and radiator is ∼3 kg/kW(e). Two SUSEE reactor options are described, based on the existing Zr/O2 cermet and the UH3/ZrH2 TRIGA nuclear fuels.

At the “Radiochemistry Laboratory” of Accelerators and Applied Superconductivity Laboratory, LASA, several production and quality assurance methods for short-lived and high specific activity radionuclides, have been developed. Presently, the irradiations are carried out at the Scanditronix MC40 cyclotron (K=38; p, d, He-4 and He-3) of JRC-Ispra, Italy, of the European Community, while both chemical purity and specific activity determination are carried out at the TRIGA MARK II research reactor of University of Pavia and at LASA itself. In order to optimize the irradiation conditions for platinum radiotracer production, both thin- and thick-target excitation function of natOs(α,xn) nuclear reactions were measured. A very selective radiochemical separation to obtain Pt radiotracers in No Carrier Added form, has been developed. Both real specific activity and chemical purity of radiotracer, have been determined by neutron activation analysis and atomic absorption spectrometry. An Isotopic Dilution Factor (ID...

TFE Verification Program is in the sixth year of a program to demonstrate the performance and lifetime of thermionic fuel elements for high power space applications. It is jointly funded by SIDO and DOE. Data from accelerated tests in FFTF and EBR‐II show component lifetimes longer than 7 years. Alumina insulators have shown good performance at high fast fluence. Graphite‐cesium reservoirs based on isotropic graphite also meet requirements. Three TFEs are current operating in the TRIGA reactor, the oldest having accumulated 15,000 hours of irradiation as of 1 October 1990.