Search

This paper presents the first determination of the spatial resolution of the ANET Compact Neutron Collimator, obtained with a measuring campaign at the LENA Mark-II TRIGA reactor in Pavia. This novel collimator consists of a sequence of collimating and absorbing channels organised in a chessboard-like geometry. It has a scalable structure both in length and in the field of view. It is characterized by an elevated collimation power within a limited length. Its scalability and compactness are added values with respect to traditional collimating system. The prototype tested in this article is composed of 4 concatenated stages, each 100mm long, with a channel width of 2.5mm, delivering a nominal L/D factor of 160. This measuring campaign illustrates the use of the ANET collimator and its potential application in neutron imaging for facilities with small or medium size neutron sources., Comment: 9 pages, 8 figures, prepared for submission to JINST

At present, the power tracking performance of nuclear Reaktor TRIGA PUSPATI (RTP) is considered unsatisfactory performance due to relatively long settling time during transient and a chattering noise during steady-state power output. Application of the conventional Feedback Control Algorithm (FCA) as a power control technique is proven to be inadequate to keep the core power output stable and within tight multiple parameter constraints for the safety demand of the RTP. Hence, the present study proposed a multipronged core power control strategy improvement through manipulation of the current Control Rod Selection Algorithm (CRSA), Control Rod Speed Design (CRSD), and Power Change Rate Constraint (PCRC) which are part of the core power control design. In this paper, the profiling and analysis of the multipronged core power control strategy are presented. The model for core power control consists of mathematical models of the reactor core, FCA controller, and a series of multipronged models. The mathematical models of the reactor core are based on the point kinetics model, thermal-hydraulic model, and reactivity model. The reactor model is integrated with the FCA controller and a combination of CRSA-CRSD-PCRC models. The power tracking performance of the proposed control strategy and conventional FCA is compared via computer simulation. Overall, the results show the multipronged FCA offers a wider options for optimum operation of the TRIGA reactor.

The PUSPATI TRIGA Mark II Research Reactor is currently been used in various applications such as Neutron Activation Analysis (NAA), Delayed Neutron Activation Analysis (DNAA), radioisotope production for medical, industrial and agricultural purposes, Neutron Radiography and Small Angle Neutron Scattering (SANS). However, there are not much research activities on utilizing neutron flux from the reactor to study properties of superconducting materials. For superconductors to be fabricated into components of electrical and magnetic devices, the sustainability of the transition temperature, TC and the transport critical current density, JC of the materials is vital to ensure the durability and effectiveness of the devices. A strong superconducting flux pinning capability of superconductors is necessary to maintain high TC and larger JC. Meanwhile, exposure of superconductors to neutron flux is able to create the self-organization of defects that are responsible in producing higher TC and JC in superconductors where the defects act as the flux pinning centres. In this paper, we report the utilization of neutron flux from the PUSPATI TRIGA Mark II research reactor in enhancing the TC and JC of bismuth-strontium-calcium-copper oxide (BSCCO) superconductor. Appropriate sample holder and shielding was fabricated as protection from radiation effect. In this work, 2212-phase BSCCO superconductor (Bi-2212 phase) samples were synthesized using the conventional solid-state reaction method. The samples underwent fast neutron irradiation with neutron flux of 2.0 × 1011 neutrons/(cm2.s) for 6 hours. Results showed that the TC of the samples improved slightly, and the JC was found to increase by more than 3 folds.

This paper focuses on the design upgrade of existing scanner of gantry radial arm robotic system for corrosion inspection on wall reactor of a TRIGA tank. The scanner design will be used for under water inspection environment verification. This inspection can be performed by using ultrasonic C-scan that will be designed in this project. This gantry radial arm system is a modular design that allows upgrading, easy dimensional matching and, specifications related to kinematic and technique. The gantry radial arm system is expected to be user-friendly. SolidWorks software was used to design the prototype of the gantry radial arm. Most of the components are made from Aluminium which is much sturdy and low cost. Meanwhile for the rod holder system, spring mechanism was used to prevent rod from bending when it moves down into the water. The test results show that the system designed is safe to be used, as it has been proved by simulation and calculation.

Characterization neutron and gamma beams at the tangential beam port of TRIGA 2000 reactor was done in order to prepare the development of Prompt Gamma Neutron Activation Analysis (PGNAA) facility. Characterization was performed by simulation using Monte Carlo method with MCNPX and PHITS computer codes. The MCNPX code was used to simulate the neutron-gamma fluence and energy spectra in the reactor core from fission reactions. The PHITS code was used to simulate the distribution of the neutron-gamma fluence and energy spectra in the tangential beam port. The simulation was done by modeling the geometry and elemental composition of reactor material and radiation source model in the form of fission reaction in the reactor core. The results indicate that the tangential beam port provides better neutron spectrum and neutron-gamma fluence ratio for PGNAA. These results will be used to simulate the colimator design and neutron interaction with samples in PGNAA facility.

Demineralized water is commonly used as cooling water, neutron moderator and as a radiation shield in a nuclear reactor such as Bandung Triga 2000 Reactor. According to its function, demineralized water has to fulfilled some requirements as stated in Safety Analysis Report (SAR) of Bandung Triga 2000 Reactor. The requirements were impurities of Ca, Na and Mg, pH and conductivity electricity in demineralized water reactor. Determination of Ca, Na and Mg as impurities were carried out using Flame Atomic Absorption Spectrophotometry (FAAS). In this activity, instrument and method verification were carried out as quality control. Instrument verification determined were sensitivity, precision and detection limit of FAAS. Method verification was carried out by using spike demineralized water with Na, Ca, and Mg standard. Accuracy and precision of the method were in the acceptance criteria according to AOAC then implemented to determined impurities of Ca, NA, Mg in demineralized water reactor samples. Na and Mg in demineralized water using FAAS gave values less than 1 mg/L, pH values were in the range 5,5- 6,5 meanwhile conductivity electricity gave values were less than 3 µmho fulfilled the requirements in SAR. These results were indispensable to support the performance of Bandung Triga 2000 Reactor which needed by public and university for the development of science and its application in the industry, health, environment and other fields.

For more than 35 years of operation and utilization of the neutron source from the PUSPATI TRIGA Reactor (RTP), the reactor has faced ageing challenged and degradation in several structures, systems and components (SSCs). Several modifications have been made since 2010 to ensure safe operation and sustainability, however, due to recent intensively request from the user to use the neutron source, it has shown degradation behaviour in the reactor SSCs including the biological shielding of the RTP. In this paper, it described the assessment of the safety barriers integrity specifically at the biological shielding concrete structure of the RTP. In-service inspection and X-Ray diffraction methodology were used to investigate and analyses the condition. The investigation results show no irregularity except slightly degradation occurrences has been noticed and aware.

TRIGA Mark II has served as nuclear research reactor for decades, which is to provide research facility towards nuclear development as well as for trainings in the aspect of education and industry. Standard TRIGA fuel, UZrH1.6 is used in this reactor which can be found in two types; aluminium or stainless steel cladding. In this study, stainless steel cladded fuel used in Reactor TRIGA PUSPATI (RTP) that consists of fissile U-235 with 19.9% enrichment and 20% weight of Uranium is used as reference for input data. The irradiated fuel inventories and fuel burnup were determined using ORIGEN2 code by referring to PWR data library, consists of gamma and neutron source term comes from activation products, actinides and fission products after 18 years of irradiation and 5 years of cooling period with 200 days of time step. The outcomes from this study can be concluded with the inventories from fission products to be the highest in all categories, in which the U-235 depleted up to 52.9%. The results from this simulation is crucial as it is then carried forward for radiation shielding for interim storage facility.

An internal fire analysis Level-I probabilistic safety assessment (PSA) for TRIGA PUSPATI (RTP) nuclear research reactor was initiated. Previous study on half-scope mode Level-I PSA, focusing only on internal initiating event, were developed for both full power and shutdown operational mode. Hence, in ensuring PSA remains relevant, a subsequent study to kept PSA up to date is a compulsory. The aim of this study is to present the step-by-step work done and the insights gained in identifying and screening significant compartments that are potential to had the worst consequences as an ignition sources. This preliminary study will only focus on RTP’s basement which consist of 3 compartments: switch room, battery room and basement hall. Procedure in data collection were based on IAEA document, Specific Safety Guide No. 3: Development and Application of Level 1 Probabilistic Safety Assessment for Nuclear Power Plants (SSG-3). Microsoft Excel was used as data collection media and internal fire hazard database development. A brief qualitative analysis using failure mode and effect analysis is also presented.

At present, Reactor TRIGA PUSPATI (RTP) is the only nuclear research reactor in operation in Malaysia since 1982. Each year, the reactor is operated on average for about 600 hours mainly for the advanced neutron and gamma radiation studies as well as for isotope production, sample activation and student training. About one-third of the RTP’s core volume is occupied by water. The water-cooling and purification systems maintain low water conductivity, remove impurities, maintained the optical clarity of the water and dissipating the reactor heat. They consist of a water surface skimmer, pump, filter, demineralizer, heat exchanger unit, associated piping and valves as well as other miscellaneous instrumentations. Thus, the aim of this study is to investigate water quality and maintains the water chemistry to the required specifications at all times in order to ensure longevity and performance of the reactor system.

The present work discusses two different models of thorium irradiation in the TRIGA 2000 reactor core to compare the neutronic characteristics of irradiation of thorium dioxide (ThO2) and thorium zirconium hydride (ThZrH). The Monte Carlo N-Particle (MCNP) Transport code, based on the Monte Carlo method, is used to design three dimensional models for TRIGA reactor core at typical operating power 2 MW. These models are used to determine the effective multiplication factor (keff) and neutron flux distribution. The ORIGEN code is used to calculate the 233U isotope production. It is found that decreasing of keff from the effect of hydride irradiation rod is a little lower than the oxide irradiation rod.

The TRIGA 2000 reactor is a nuclear installation which is operated at 1000 kW. Gamma and neutron radiation exposure will contribute to occupational doses that will be received by operators and other radiation workers. To maintain the safety of radiation workers, assessment of gamma and neutron radiation is needed in the TRIGA 2000 reactor. survey in the hall reactor using the RTD 30 and RadEye NL detectors. In addition to the direct measurement, calculations are also carried out with simulation data dose using Monte Carlo MCNPX for the calculation and followed by PHITS to get the graphical dose distribution of gamma and neutron. The study will focus on the assessment of the diffusion Gamma and neutron doses by doing the measurement and simulation data collect for the aim of the optimization of the Radiation protection policy in the reactor TRIGA 2000 Bandung. The result of the measurement is shown with graph in different power taken during the measurement and the simulation is the graphical distribution dose of gamma and neutron using PHITS code. A preventive dose reduction technique, worker and publics protection elaborate to optimise the Radiation protection program.

The TRIGA 2000 research reactor has been operating since mid 2000. Currently, there are plans to convert the reactor core from cylindrical fuel element to plate type fuel element. The conversion process is expected to not change too many existing systems, including the primary cooling systems. For this reason, it is necessary to evaluate the pump capability of the primary cooling system. Based on experimental studies, then merging with the existing installation system head, it has been obtained the main and the redundant pump mass flow rate of 45 and 45.5 kg/s, respectively. The pump capability decreased by around 5 % compared to the results of its commissioning. If the two pumps are combined in parallel, the combined mass flow rate can increase to 62 kg/s. The existing primary pumps cannot be used for primary cooling system if the reactor core is converted into a plate type, even though both pump are installed in parallel. These results need to be considered in designing a new core of the Bandung TRIGA 2000 research reactor with primary cooling systems that are slightly modified because they have to add the decay tank.

The development of the Boron Neutron Capture Therapy (BNCT) facility in Malaysia can be performed at the thermal column of the Malaysia research reactor. TRIGA MARK II is one of the facilities that can provide neutron source for BNCT facility. The specification of neutron flux and the gamma dose rate must consider for the development of the BNCT facility as a safety precaution for this research. Based on previous research, the thermal column identified as a suitable place for BNCT facility. To design the neutron collimator for BNCT purpose, the characterization of material towards thermal neutron flux explored using TLD and Microspec-6 and the collimator design was simulate using Monte Carlo N-Particle (MCNP) software based on the characterize materials in order to produce high thermal neutron flux. The combination of lead, HDPE, 30% borated polyethylene and aluminium as collimator design D1 simulate the highest thermal neutron 1.5770 x 109 neutron.cm−2s−1 and suitable for BNCT research purpose at the thermal column.

Analysis of the TRIGA 2000 core reshuffling scenario based on fuel burn-up and fuel density has been done. One of the requirements of Bandung TRIGA 2000 Reactor reliability is the implementation of maintenance and surveillance activities with predetermined Operation and Limits Conditions (OLC) so that the safety parameters will be met. Core management is one of the maintenance activities carried out. Safety parameters that are required to be fulfilled in the core management include the shutdown margin and the power peaking factors. The scenario carried out in the core management in this analysis that was doing the core compaction by sorting the fuel position on the core based on the fuel burnup and the fuel density. Each core management method was carried out in 4 scenarios. The purpose of this analysis is of course to compare both methods and also to get the optimal scenario to be applied to the core of the existing Bandung TRIGA 2000 Reactor. From the eight proposed scenarios, obtained two scenarios that meet the requirements of shutdown margin and power peaking factor from the OLC, namely scenario 2 with the k-eff value obtained 1.01243, and scenario D with the k-eff value obtained 1.02031. Therefore, scenarios 2 and D will then be proposed as the existing Bandung TRIGA 2000 Reactor core configuration scenario.

Iodine in food usually presents in low-level concentration, so it is not easy for measurement by thermal NAA related to the high background of the Compton. Therefore, the aim of this work is to develop the iodine analysis method in a food matrix, so that the iodine analysis method is obtained, which is guaranteed the validity. In this activity, a number of 50 mg of SRM NIST 1548a Typical Diet, IPE sample and standard were irradiated in a pneumatic system facility at TRIGA 2000 reactor with the power of 700 KW by covering sample using cadmium plates to absorb thermal neutrons. The results of the iodine-128 spectrum analysis at 442.3 Kev in the GENIE 2000 software showed a good prospect for iodine detection, with a limit of detection (LOD) is 0.19 mg/kg. The results of the validation method by accuracy and precision testing stated with % Recovery and RSDr each ranged between 93.6-103.1% and 16.33-16.60, respectively. This data have good agreement with accuracy and precision. Criteria the preliminary study, it can be concluded that iodine at food matrices could be determined using epithermal NAA.

The continuity of the TRIGA 2000 reactor operation is determined by the fuel temperature and the primary cooling water temperature. Currently operating the TRIGA 2000 reactor using 105 fuel configurations it is difficult to reach 1000 kW of power, because the fuel temperature and primary coolant temperature in the core are high. It causes boiling and bubble formation in the reactor core, thus reducing neutron moderation in the core. The results of neutronic calculations on the configuration of 105 fuels in the core at this time, it is known that heat generation is focused into the center of the core, and some fuels have large power fluxes, so that some of these fuels produce high cladding surface temperatures as well and cause boiling. In this research, thermal hydraulic analysis has been carried out using the CFD program package for the configuration of 105 fuels in the core. Based on the results of research on the reactor operated at 500 kW power, the hottest channel is at B5 fuel with the power it generates of 9.43 kW, the maximum surface temperature of the fuel cladding is 126.41 °C, and the temperature of coolant wets 119.92 °C. This temperature has exceeded the saturation temperature of the reactor cooling water 112.4 °C, so that sub-cooled boiling is possible. This situation is in accordance with the conditions that occur in the operation of the TRIGA 2000 reactor, where at 500 kW power has begun to observe bubbles coming out of the reactor core. One effort that can be done to reduce the temperature of the TRIGA 2000 reactor core is to do reshuffling and increase the amount of fuel used.

The thermal column at the TRIGA PUSPATI (RTP) research reactor can produce thermal neutron. However, the optimization on the thermal neutron flux produced should be performed to gain a sufficient thermal neutron for boron neutron capture therapy purpose. Thus, the objective of this review is to optimize the thermal neutron flux by designing the collimator with different materials at the thermal column. In order to fulfil the requirement, set by the IAEA standard, the study of BNCT around the world was being reviewed to study the suitable measurement, material, design, and modification for BNCT at the thermal column of TRIGA MARK-II, Malaysia. Initially, the BNCT mechanisms and history was review. Then, this paper review on the design and modifications for BNCT purpose around the world. Based on this review, suitable material and design can be used for the BNCT in Malaysia. Moreover, this paper also reviews the current status of BNCT at the RTP with the measurement of the thermal neutron flux was conducted along the thermal column at 250 kW. The thermal column of RTP was divided into 3 phases (Phase 1, Phase 2 and Phase 3) so that an accurate measurement can be obtained by using gold foil activation method. This value was used as a benchmark for the neutron flux produced from the thermal column. The collimator was designed using different types of materials, and their characteristic towards gamma and neutron flux was investigated. The reviewed demonstrated that the final thermal neutron flux produced was significantly for BNCT purpose. Lastly, this paper recommends the future research can be conducted on BNCT at RTP.

Neutron Activation Analysis (NAA) is a nuclear analytical technique for elemental analysis that has been widely applied in various fields such as environment, industry, health and food. It has high specificity and sensitivity for the analysis of elements with very low concentrations up to the picogram order. In Indonesia, NAA technique was developed in BATAN that has a research reactor as a neutron source. Relative method of NAA that uses the multi-elements standard as a comparison is widely used in Indonesia. At present, BATAN is developing an absolute method k0-NAA that has several advantages such as faster and more economics due to the reduction of multi-elements standard dependence. BATAN k0 Utility (BAkUL) beta version software has also been developed for data processing of the absolute methods. k0-NAA requires reactor parameters (α and f) as well as detector parameters for data analysis. The parameter f is the ratio of the thermal neutron flux to the epithermal whereas α is the epithermal spectrum deviation parameter from the 1/E distribution. The aim of this study is to measure parameters a and f using the triple bare monitor method by irradiating simultaneously Au and Zr without Cd cover. Measurement of reactor TRIGA 2000 parameters (α and f) were carried out at the irradiation positions CT1-1 and G17-2, and the results for α and f parameters at the CT1-1 irradiation position were -0,100 and 8.22 respectively, while at the G17-2 position was - 0.087 and 13.8. Each irradiation position has different α and f values so it is necessary to determine the reactor parameters for other irradiation positions. The measurement results α and f can be used for the next samples irradiation and as a database to support the application of the k0-NAA method in Center for Applied Nuclear Science and Technology (PSTNT) BATAN.

A conversion of the TRIGA reactor Bandung has been planned by replacing fuel from rod type to plate type. The fuel replacement causes the fuel arrangement or reactor core configuration should be changed as well as the cooling system. Currently, the reactor cooling system takes place by natural convection with the flow direction from the bottom to the top of the core. The replacement of plate-type fuel with a small distance between plates causes the core cooling process to decrease, therefore to make more effective the forced convection cooling system with the direction of upward flow is required. It is expected that the cooling process will increase. The ability of the cooling system depends on flow distribution in the core at the time of reactor in operating, therefore it is necessary to analyze the distribution of coolant flow and the velocity distribution of the cooling system from the top reactor pool to the reactor core to determine that the coolant flow from the inlet pipe mostly goes to the core. The purpose of this research is to analyze the distribution of coolant flow and flow velocity to the reactor core. The analysis was carried out in a simulation using computational fluid dynamics FLUENT software by creating reactor core modeling. The result shows that the cooling water flows from above along the reactor pool, then the stored energy is reduced so that the flowing slowdown. In this case, resulting in evenly flow distribution to the reactor core.

Boron Neutron Capture Therapy (BNCT) is a nuclear application to cure cancer by using the interaction between the boron compound and the slow (thermal and epithermal) neutron beam obtained from the neutron source. The previous feasibility study shows that the thermal column can generate higher thermal neutrons for BNCT application at the TRIGA MARK II reactor. Currently, the BNCT facility is planned to be developed at the thermal column. Thus, to fabricate the nuclear application such as BNCT required several required such as the study of design, material, neutron desired and safety for the application. Therefore, the use of simulation software such as Monte Carlo N-Particle (MCNP) was extensively essential to identify and determine the most convenient design of the BNCT facility with proper safety requirement standard. Besides that, the simulation also useful in terms of identifying good material and best geometry and the uses of simulation can guide to make sure this research achieves the best results. Generally, MCNP function to track the particle over its energy range using the irradiation transport and the neutron flux and the gamma flux produced from the simulation using MCNPX code was consider right if the relative errors contributed was lower than 5%. This research simulates that the uncovered collimator provides high neutron flux and high collimator performance compared to the fully covered collimator. This research is useful in the simulate on the desired neutron from the BNCT facility design at the TRIGA MARK II reactor.

Reflector TRIGA Mark II Bandung waste container shielding using Microshield 7.02 has been conducted. In 1996, TRIGA Mark II power was upgraded from 1MWt into 2MWt, and the reflector was one part of the reactor core that was changed. Reflector waste consists of graphite with activated Co-60. In 2015 the activity was 0.557 Ci with 1,221.5 Kg weight. This large radioactive waste needs to be removed from Bandung Nuclear Area into PTLR-Serpong Nuclear Area. This removal process needs calculations and the basic design of the waste container. Pure lead material is planned to be used for attaining requirements standards of radioactive container category base on Transport Index score. In addition, MicroShield 7.02 is a software that is used for helping to analyze shield thickness and could count exposure rate out of the waste container. This software was also used in this research to calculate the lead thickness and demonstrate the radioactivity in several points. The result of simulation with this software is; lead thickness for the container is 3 cm; the container category is III-Yellow with; Transport Index score is 189 or bigger than 10. With tubular shape with 3 cm thickness it will use 1,945.13 kg pure lead. Then, the dose rate on the surface of the container is 128.6 mSv/hr.

This research offers a fuel conversion from cylinder type to plate type for solving the operation problems as a solution when the TRIGA reactor fuel production run out. The research has been implemented utilizing MCNP6 software to model and simulate the core of the TRIGA 2000 reactor plate type fuel. The geometry of the simulated reactor consists of a reflector, a moderator and a reactor tank. The fuel was U3Si2Al with 19.75% enrichment and the density was 2.96 gU/cm3, the reflector was graphite and the control rod was AgInCd. The first criticality value with 15 fuel assemblies and the 4 control rods position at 11.97cm from the top surface of the reactor core was (1.00000 ± 0.00016) in agreement with the reference value for initial configuration of TRIGA reactor core was (1.00 ± 0.01).

The fundamental advantage in using Monte Carlo methods for burnup calculations is to formulate an effective optimal fuel management strategy for the TRR-1/M1 research reactor. The core management study has been performed by utilizing the essentially parameters including multiplication factor, power peaking, neutron flux and burnup calculation based on the Monte Carlo calculation. The fuel element burnup was calculated after reshuffling the reactor core. The fuel cycle length and core parameters such as core excess reactivity, neutron flux, axial and radial power factors and other parameters are determined. The core excess reactivity was calculated as a function of burnup. The maximum excess reactivity shall not exceed 6.3% Δk/k. The maximum fuel temperature shall not exceed 930 °C during steady-state operation. Typically, a core loading operated with the maximum burnup between 100 to 200 MWD depending on the utilization requirements. The thermal neutron flux in the irradiation positions is within the order of 1011 - 1013 n/cm2-sec. The study gives valuable results into the behaviour of the TRR-1/M1 research reactor and will ensure optimized utilization and operation of the reactor during its life time. It will establish the strategic planning for fuel management in the reshuffling and reloading schemes patterns and its safe implementation in the future.

Two series of measurements were performed in the JSI TRIGA research reactor in 2014 and 2017 to validate the 55Mn(n,γ)56Mn cross-sections and experimentally investigate the relationship between the 55Mn(n,γ)56Mn reaction and the rate of tritium production through the 6Li(n,t)4He reaction. Indeed, previously observed similarities between the sensitivity profiles of the neutron reaction of tritium production on lithium, 6Li(n,t)4He, and those of the 55Mn(n,γ)56Mn reaction in tritium breeder modules indicated that the latter reaction could be used as an effective monitor of tritium production, at least for short-term monitoring (the half-life of 56Mn being 2.579 h). However, experimental verification, improvements and validation of the 55Mn(n,γ)56Mn cross-sections are needed in order to meet the required accuracy. Foils of certified reference material Al–1%Mn, as well as LiF thermoluminescent detectors and Li2O samples were irradiated, both bare and under cadmium, to study the potential use of the 55Mn(n,γ)56Mn reaction for monitoring tritium production in fusion devices. Additionally, Al–0.1%Au was also irradiated for comparison, the 197Au(n,γ)198Au reaction cross-section being a standard. In order to obtain complementary information for data validation purposes, the irradiations were performed in positions within the JSI TRIGA reactor with different neutron spectra, i.e. in the central channel, the pneumatic tube and the F19 position, both in the outer ‘F’ ring of the reactor core and in the IC-40 irradiation channel located in the graphite reflector surrounding the reactor core. Bare and cadmium-covered irradiations were needed to subtract the contribution of epithermal neutrons to the 55Mn(n,γ)56Mn reaction. Calculations of the reaction rates were performed using the Monte Carlo code MCNP6.1 with a detailed model of the JSI TRIGA reactor, with the samples, the irradiation capsules and covers being modelled explicitly. The uncertainties involved in the measurements and the calculations were carefully evaluated. The principal objective was to study the energy response and correlations between the 55Mn(n,γ)56Mn reaction in irradiated Al–1%Mn and the 6Li(n,t)4He reaction in irradiated LiF and Li2O. Good consistency between the measured and calculated 55Mn(n,γ)56Mn and 197Au(n,γ)198Au reaction rates, in most cases within the uncertainty bars, was observed.

Spent fuel pool uses water to temporarily cool spent nuclear fuels before final disposal. The Malaysian Nuclear Agency is recently working on its spent fuel pool facility to store irradiated TRIGA fuel rods. These irradiated fuels still contain significant amount of fissile material which are capable to induce criticality in the fuel storage facility. This study is therefore performed to investigate the safety criticality analysis for the proposed fuel storage rack design of 5x5 arrays. MCNP5 code was used to model the 5x5 array of fuel storage rack in order to determine the optimal fuel pitch distance. The analysis was based on the standard UZrH1.6 TRIGA fuel cladded with stainless steel used in PUSPATI TRIGA Reactor, which is 19.9% enriched 235U and 20% weight of Uranium. By varying the pitch distance between 4 cm to 14 cm with no neutron absorber included, the optimum fuel rack design was determined where sub-criticality condition can be maintained.

Characterization of neutron beam compositions is important in order to capture great radiographic images. The quality of the neutron beam has significant impacts to the radiographic images produced using that beam. Characterizing of the beam, including determination of the effective value for the thermal neutron content, scattered neutron content, gamma content, and pair production content as well as beam sensitivity, is vital for producing quality radiographic images. This paper provides a characterization of the neutron radiography’s beam port (beam port 3) at TRIGA Mark II PUSPATI research reactor (RTP). The experiments to determine the effective beam contents and neutron radiograph sensitivity at the beam are based on American Society for Testing and Materials (ASTM) standards document E545. The Beam Purity Indicator (BPI) and Sensitivity Indicator (SI) are the standard used in these experiments. Neutron radiographic imaging of this standard was performed using direct radiographic method with 0.1 mm thickness Gadolinium (Gd) metal converter and film. Characterization of the beam was carried out for 10, 15, 20, 25 and 30 minutes exposure time at 750 kW reactor operating power. The standard was placed at 110 cm from the beam opening. The current neutron beam performances are compared with previous results.

Fuel Temperature reactivity coefficient is one of the important operational and safety parameters in nuclear reactors to monitor for normal operation and transient safety analysis in research as well as in power reactors. The aim of this work is to estimate the temperature reactivity coefficient by using adaptive neuro-fuzzy inference system (ANFIS) as a preliminary assessment for developing an online monitoring system in research reactors such as the Reactor TRIGA PUSPATI (RTP). The input for the learning mechanism is the fuel temperature recorded at Ch A of the instrumented fuel element (IFE) and control rod positions with the fuel temperature reactivity coefficient as the targeted output. The ANFIS model was validated using testing data that consisted of fuel temperature measured at Ch B and results on -0.00571 $°C−1 (-3.997 pcm°C−1) of actual output and -0.00592 $°C−1 (4.144 pcm°C−1) of predicted output. Through the developed model, it showed that the ANFIS also can be used as a tool to estimate the fuel temperature reactivity coefficient with good prediction and accuracy shown from the calculation of mean absolute error (MAE) of 0.00349.

This paper discusses the overview of the development of TRIMON, a next-generation core management code for TRIGA reactors. The newly developed code employs Monte Carlo method which uses diffusion theory type homogenized neutron cross section data, thus, speeding up the stochastic simulation via simplification of a complex core geometry. It has the capability of running three-dimensional core simulation up to 16 neutron energy groups. In general, the group neutron cross sections are pre-processed and homogenized via the Effective Diffusion Homogenization (EDH) method. The important features of TRIMON include core eigenvalue calculation, fuel burnup management, three-dimensional power and neutron flux distribution prediction. In this paper, the numerical results obtained using TRIMON were compared with the actual measured operational data. The performance comparison was also made between TRIMON and MCNP.

Currently plate-type fuels are planned to use in the research reactor TRIGA-2000 Bandung, Indonesia (further called as TRIGA-Plate reactor). Plate type fuel consists of enriched uranium sandwiched between metal cladding. Plate-type fuel is used in some research reactors to obtain high neutron flux and is used to study material irradiation or isotope production. For the purpose of Safety Analysis Report (SAR) on the fuel-type modification plan of TRIGA-2000 Bandung, it is necessary to calculate radioactivity inventory from the reactor core under operating conditions. The radioactivity inventory of materials irradiated in a reactor depends on the initial material composition at the BOC, burn-up history from BOC to EOC and the power peaking factor. The purpose of the present work is to evaluate the fuel burn-up and radioactivity inventory in fuel materials of TRIGA-Plate research reactor. The mass of U-235 consumed and the mass of Pu-239 and Th-232 produced are also evaluated. The calculation results obtained using ORIGEN2 code are: the mass of U-235 consumed is 1.40E+02 gram; while the mass of Pu-239 produced is 1.36E+00 gram and Th-232 produced is 1.03E-06 gram. The largest radioactivity inventories of the reactor at the EOC sequentially are: Kr group is about 2.29E+02 Ci; for group I is 7.77E+03 Ci and for groups Cs is 1.92E+02Ci.

The fuel of TRIGA core is specific and raise a big problem for a continuous operation of reactor. It was no longer produced by TRIGA International anymore. Meanwhile, the demand of radioisotope production in Indonesia and also in the world is still increasing. Therefore, the research reactor which has irradiation facilities and neutron flux for radioisotope production still needed. The TRIGA MARK II core conversion to plate-type fuel can be an appropriate solution. The fuel that used for conceptual design TRIGA MARK II core is MTR type, low enriched uranium, 19.75% enriched in 235U. It is important to study the effect of reflector on neutronic parameter to determine the optimal neutron flux in the core. Optimization of neutron fluxes in irradiation positions is one of great concern in a research reactor utilization. The approach in this research is based upon an optimization of the core configuration combined with improvement of reflector characteristics. This simulation study conducted by using the standard reactor physics simulation codes WIMSD-5B and Batan-FUEL. The present study is to analyze the effect of different reflector on neutronic parameter using diffusion code. TRIGA core design comprise 16 fuel elements and 4 control rods put in 5 x 5 position of grids plate and by loading reflector elements at outside the core. The aim of this research is to get the optimal design such as the highest neutron fluxes at the central flux trap position and also at irradiation positions in the core. it also becomes fuel cycle length of this core a concern. The requirement for the neutron flux at trap irradiation position is not less than 6x1013 n/cm2s at 2 MW of power. The core configuration which proposed meets the acceptance criteria such as stuck rod condition, thermal neutron flux at the trap position and shutdown margin. But it still has the lower burn up criteria for its safe operation at 2 MW of power.

bstract The conversion of reactor fuel type will affect the heat transfer process resulting from the reactor core to the cooling system. This conversion resulted in changes to the cooling system performance and parameters of operation and design of key components of the reactor coolant system, especially the primary cooling system. The calculation of the operating parameters of the primary cooling system of the reactor TRIGA 2000 Bandung is done using ChemCad Package 6.1.4. The calculation of the operating parameters of the cooling system is based on mass and energy balance in each coolant flow path and unit components. Output calculation is the temperature, pressure and flow rate of the coolant used in the cooling process. The results of a simulation of the performance of the primary cooling system indicate that if the primary cooling system operates with a single pump or coolant mass flow rate of 60 kg/s, it will obtain the reactor inlet and outlet temperature respectively 32.2 °C and 40.2 °C. But if it operates with two pumps with a capacity of 75% or coolant mass flow rate of 90 kg/s, the obtained reactor inlet, and outlet temperature respectively 32.9 °C and 38.2 °C. Both models are qualified as a primary coolant for the primary coolant temperature is still below the permitted limit is 49.0 °C.

TRIGA 2000 reactor in Bandung is planned to have the fuel element replaced, from cylindrical uranium and zirconium-hydride (U-ZrH) alloy to U3Si2-Al plate type of low enriched uranium of 19.75% with uranium density of 2.96 gU/cm3, while the reactor power is maintained at 2 MW. This change is planned to anticipate the discontinuity of TRIGA fuel element production. The selection of this plate-type fuel element is supported by the fact that such fuel type has been produced in Indonesia and used in MPR-30 safely since 2000. The core configuration of plate-type-fuelled TRIGA reactor requires coolant flow rate through each fuel element channel in order to meet its safety function. This paper is aimed to describe the results of coolant flow rate distribution in the TRIGA core that meets the safety function at normal operation condition, physical test, shutdown, and at initial event of loss of coolant flow due power supply interruption. The design analysis to determine coolant flow rate in this paper employs CAUDVAP and COOLODN computation code. The designed coolant flow rate that meets the safety criteria of departure from nucleate boiling ratio (DNBR), onset of flow instability ratio (OFIR), and ΔΤ onset of nucleate boiling (ONB), indicates that the minimum flow rate required to cool the plate-type fuelled TRIGA core at 2 MW is 80 kg/s. Therefore, it can be concluded that the operating limitation condition (OLC) for the minimum flow rate is 80 kg/s; the 72 kg/s is to cool the active core; while the minimum flow rate for coolant flow rate drop is limited to 68 kg/s with the coolant inlet temperature 35°C. This thermohydraulic design also provides cooling for 4 positions irradiation position (IP) utilization and 1 central irradiation position (CIP) with end fitting inner diameter (ID) of 10 mm and 20 mm, respectively.

The thermal neutron flux measurement has been conducted at the out-core location using self-powered neutron detectors (SPNDs). This work represents the first attempt to study SPNDs as neutron flux sensor for developing the fault detection system (FDS) focusing on neutron flux parameters. The study was conducted to test the reliability of the SPND's signal by measuring the neutron flux through the interaction between neutrons and emitter materials of the SPNDs. Three SPNDs were used to measure the flux at four different radial locations which located at the fission chamber cylinder, 10cm above graphite reflector, between graphite reflector and tank liner and fuel rack. The measurements were conducted at 750 kW reactor power. The outputs from SPNDs were collected through data acquisition system and were corrected to obtain the actual neutron flux due to delayed responses from SPNDs. The measurements showed that thermal neutron flux between fission chamber location near to the tank liner and fuel rack were between 5.18 × 1011 nv to 8.45 × 109 nv. The average thermal neutron flux showed a good agreement with those from previous studies that has been made using simulation at the same core configuration at the nearest irradiation facilities with detector locations.

Neutron and gamma ray fluences are the important criteria to be looked at in a neutron diffractometer system facility design. The fluence is defined as the number of neutrons and gamma rays travel through a unit area. Currently, the facility for neutron diffractometer system are planned to be developed at radial beam port 1 of TRIGA MARK II PUSPATI research reactor (RTP). The aim of this research is to determine the value of neutron and gamma ray fluences produced at the end of radial beam port 1 and to identify the shielding materials suitable for neutron and gamma ray. In order to achieve this aim, an experiment has been designed to obtain the neutron and gamma dose rates by using TLD-600 and TLD-700. The results from this experiments are converted into neutron and gamma ray fluences and are then compared with the results from simulation. The comparison shows that both results meet an agreement on the feasibility of shielding material for neutron diffractometer system. Our research results may be of help in the design of shielding material for neutron diffractometer facility at RTP.

Measurements of in-phantom dose distributions and images are important for Boron Neutron Capture Therapy treatment planning. The method for spatial determination of absorbed doses in thermal or epithermal neutron fields, based on Fricke-xylenol-orange-infused gel dosimeters in form of layers, has revealed to be very reliable, as gel layer dosimeters give the possibility of obtaining spatial dose distributions and measurements of each dose contribution in neutron fields, by means of a properly studied procedure. Quite recently, BNCT has been applied to treat liver metastases; in this work the results of in-phantom dosimetry for explanted liver in BNCT treatments are described. Moreover, polyacrylamide gel (PAG) dosimeters in which a polymerization process appears as a consequence of absorbed dose, have been recently tested, because of their characteristic absence of diffusion. In fact, due to the diffusion of ferric ions, Fricke-gel dosimeters require prompt analysis after exposure to avoid spatial information loss. In this work the preliminary results of a study about the reliability of polymer gel in BNCT dosimetry are also discussed. Gel layers have been irradiated in a phantom exposed in the thermal column of the TRIGA MARK II reactor (Pavia). The results obtained with the two kinds of gel dosimeter have been compared.

The aim of the present paper is to verify the applicability of neutron activation analysis to a metallurgic problem as a possible alternative technique to the standard investigation methodologies. A first series of measurements was performed in order to check the feasibility of irradiation and counting over metallic samples. Some of the feared problems concerned an excessive activation of the matrix and the consequent difficulties in the spectrum interpretation, as well as the removal of the radioactive waste created by the irradiation. Afterwards, a second series of measurements was performed to collect results aimed at the solution of a specific metallurgic case. The tests were performed at the TRIGA MARK II reactor facility of the LENA (Laboratorio Energia Nucleare Applicata) Institute of the Pavia University.

This paper presents the fuel burnup analysis of the Thai research reactor (TRR-1/M1), TRIGA Mark-III, operated by Thailand Institute of Nuclear Technology (TINT) in Bangkok, Thailand. The modelling software used in this analysis is MCNPX (MCNP eXtended) version 2.6.0, a Fortran90 Monte Carlo radiation transport computer code. The analysis results will cover the core excess reactivity, neutron fluxes at the irradiation positions and neutron detector tubes, power distribution, fuel burnup, and fission products based on fuel cycle of first reactor core arrangement.

The objective of this work is to demonstrate the method for validating the predication of the calculation methods for neutron flux distribution in the irradiation tubes of TRIGA research reactor (TRR-1/M1) using the MCNP computer code model. The reaction rate using in the experiment includes 27Al(n, α)24Na and 197Au(n, γ)198Au reactions. Aluminium (99.9 wt%) and gold (0.1 wt%) foils and the gold foils covered with cadmium were irradiated in 9 locations in the core referred to as CT, C8, C12, F3, F12, F22, F29, G5, and G33. The experimental results were compared to the calculations performed using MCNP which consisted of the detailed geometrical model of the reactor core. The results from the experimental and calculated normalized reaction rates in the reactor core are in good agreement for both reactions showing that the material and geometrical properties of the reactor core are modelled very well. The results indicated that the difference between the experimental measurements and the calculation of the reactor core using the MCNP geometrical model was below 10%. In conclusion the MCNP computational model which was used to calculate the neutron flux and reaction rate distribution in the reactor core can be used for others reactor core parameters including neutron spectra calculation, dose rate calculation, power peaking factors calculation and optimization of research reactor utilization in the future with the confidence in the accuracy and reliability of the calculation.

The objective of this work is to design a sample carrier for neutron irradiation experiment at beam ports of research nuclear reactor, the Reaktor TRIGA PUSPATI (RTP). The sample carrier was designed so that irradiation experiment can be performed safely by researchers. This development will resolve the transferring of sample issues faced by the researchers at the facility when performing neutron irradiation studies. The function of sample carrier is to ensure the sample for the irradiation process can be transferred into and out from the beam port of the reactor safely and effectively. The design model used was House of Quality Method (HOQ) which is usually used for developing specifications for product and develop numerical target to work towards and determining how well we can meet up to the needs. The chosen sample carrier (product) consists of cylindrical casing shape with hydraulic cylinders transportation method. The sample placing can be done manually, locomotion was by wheel while shielding used was made of boron materials. The sample carrier design can shield thermal neutron during irradiation of sample so that only low fluencies fast neutron irradiates the sample.

A liquid methane cold source has been installed in one of the tangential beam tubes of the reactor TRIGA Mark II at the 'Jozef Stefan' Institute and has given continuous and reliable service. The design and construction of the cooling system and cold source assembly are described and some graphs showing the performance are given.