Papers by Kanokrat TIYAPUN
Radiation Measurement with Data Acquisition Using LabVIEW
Current Applied Science and Technology
Radiation monitoring and measurement are important for radiation safety. Such monitoring and meas... more Radiation monitoring and measurement are important for radiation safety. Such monitoring and measurement must be carried out in order to investigate workplace conditions and individual exposures, to ensure that radiological conditions in the workplace are acceptable, safe, and satisfactory, and to keep records of radiation monitoring over a long period of time in accordance with regulations or as good practice. However, radiation monitoring instruments and system are expensive. A cost-effective radiation monitoring system using Geiger Muller (GM) radiation detector was developed. A high voltage generation circuit and data acquisition were implemented using LabVIEW software. The cost-effective radiation monitoring system is easy to use, and it can detect radiation, analyzed and record the data in a personal computer. This newly developed system is simple and can provide the radiation dose data for further analysis. An experiment was conducted to compare and calibrate the collected da...
Enhancement of Nuclear Safety in Seismic Analysis for TRR-1/M1 after Fukushima Daiichi Accident
Journal of Physics: Conference Series, 2019
The fundamental advantage in using Monte Carlo methods for burnup calculations is to formulate an... more 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 ºC 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 10 11-10 13 n/cm 2-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.
Validation of the MCNP computational model for neutron flux distribution with the neutron activation analysis measurement
Journal of Physics: Conference Series, 2015
The objective of this work is to demonstrate the method for validating the predication of the cal... more 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.
Journal of Physics: Conference Series, 2017
The neutronic analysis of TRIGA Mark II reactor has been performed. A detailed model of the react... more The neutronic analysis of TRIGA Mark II reactor has been performed. A detailed model of the reactor core was conducted including standard fuel elements, fuel follower control rods, and irradiation devices. As the approach to safety nuclear design are based on determining the criticality (keff), reactivity worth, reactivity excess, hot rod power factor and power peaking of the reactor, the MCNPX code had been used to calculate the nuclear parameters for different core configuration designs. The thermal-hydraulic model has been developed using COOLOD-N2 for steady state, using the nuclear parameters and power distribution results from MCNPX calculation. The objective of the thermal-hydraulic model is to determine the thermal safety margin and to ensure that the fuel integrity is maintained during steady state as well as during abnormal condition at full power. The hot channel fuel centerline temperature, fuel surface temperature, cladding surface temperature, the departure from nucleate boiling (DNB) and DNB ratio were determined. The good agreement between experimental data and simulation concerning reactor criticality proves the reliability of the methodology of analysis from neutronic and thermal hydraulic perspective.
Nuclear Knowledge Management
Solution for oil crisis problem
Epithermal neutron beam design at the Oregon State University TRIGA Mark II reactor (OSTR) based on Monte Carlo methods
Production of industrial and medical radioisotopes in accelerator production of tritium (APT)
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Papers by Kanokrat TIYAPUN