To non-destructively determine the burnup of a spent nuclear fuel assembly, it is essential to analyze the nuclear isotopes present in the assembly and detect the neutrons and gamma rays emitted from these isotopes. Specifically, gamma-ray measurement methods can utilize a single radiation measurement value of 137Cs or measure based on the energy peak ratio of Cs isotopes such as 134Cs/137Cs and 154Eu/137Cs. In this study, we validated the extent to which the results of gamma-ray measurements using cadmium zinc telluride (CZT) sensors based on 137Cs could be accurately simulated by implementing identical conditions on MCNP. To simulate measurement scenarios using a lead collimator, we propose equations that represent radiation behavior that reaches the detector by assuming “Direct hit” and “Penetration with attenuation” situations. The results obtained from MCNP confirmed an increase in measurement efficiency by 0.47 times when using the CZT detector, demonstrating the efficacy of the measurement system.

Concerning the apprehensions about naturally occurring radioactive materials (NORM) residues, the International Atomic Energy Agency (IAEA) and its member nations have acknowledged the imperative to ensure the radiation safety of NORM industries. Residues with elevated radioactivity concentrations are predominantly produced during NORM processing, in the form of scale and sludge, referred to as technically enhanced NORM (TENORM). Substantial quantities of TENORM residues have been released externally due to the dismantling of NORM processing factories. These residues become concentrated and fixed in scale inside scrap pipes. To assess the radioactivity of scales in pipes of various shapes, a Monte Carlo simulation was employed to determine dose rates corresponding to the action level in TENORM regulations for different pipe diameters and thicknesses. Onsite gamma spectrometry was conducted on a scrap iron pipe from the titanium dioxide manufacturing factory. The measured dose rate on the pipe enabled the estimation of NORM concentration in the pipe scale onsite. The derived action level in dose rate can be applied in the NORM regulation procedure for on-site judgments.

The nuclear fuel that melted during the Fukushima nuclear accident in 2011 is still being cooled by water. In this process, contaminated water containing radioactive substances such as cesium and strontium is generated. The total amount of radioactive pollutants released by the natural environment due to the nuclear accident in Fukushima in 2011 is estimated to be 900 PBq, of which 10 to 37 PBq for cesium. Radioactive cesium (137Cs) is a potassium analog that exists in the water in the form of cations with similar daytime behavior and a small hydration radius and is recognized as a radioactive nuclide that has the greatest impact on the environment due to its long half-life (about 30 years), high solubility and diffusion coefficient, and gamma-ray emission. In this study, alginate beads were designed using Prussian blue, known as a material that selectively adsorbs cesium for removal and detection of cesium. To confirm the adsorption performance of the produced Prussian blue, immersion experiments were conducted using Cs standard solution, and MCNP simulations were performed by modeling 1L reservoir to conduct experiments using radioactive Cs in the future. An adsorption experiment was conducted with water containing standard cesium solution using alginate beads impregnated with Prussian blue. The adsorption experiment tested how much cesium of the same concentration was adsorbed over time. As a result, it was found that Prussian blue beads removed about 80% of cesium within 10-15 minutes. In addition, MCNP simulation was performed using a 1 L reservoir and a 3inch NaI detector to optimize the amount of Prussian blue. The results of comparing the efficiency according to the Prussian volume was shown. It showed that our designed system holds great promise for the cleanup and detection of radioactive cesium contaminated seawater around nuclear plants and/or after nuclear accidents. Thus, this work is expected to provide insights into the fundamental MCNP simulation based optimization of Prussian blue for cesium removal and this work based MCNP simulation will pave the way for various practical applications.

To obtain the gamma-ray energy spectrum of artificial radionuclides which is difficult to obtain practically, virtual gamma-ray energy spectrum simulator program was developed. It can be applied for the predetermined measurement condition for which the database was developed through computational simulation and actual measurement of background radiation. For gamma spectrometry training for KHNP HPGe detectors using this program, the database for KNPG HPGe detectors was developed. First, the geometry of the detector in the simulation was adjusted to resemble the real structure by comparing the actually measured net counts rate at the main gamma peak with the value simulated by MCNP6. The Certified Reference material (CRM) of 137Cs and 60Co were used for verification. The comparison was made with respect to the situation where CRM was attached to the top and side of the detection part of the considered detector. The geometry structures of detectors were simulated by reflecting the design drawing of the products, and the simulation was performed for several thicknesses of the Ge/Li dead layer in consideration of the change in the thickness over time. As the results, the simulation geometry was tuned so that the results for 137Cs showed a difference within 10% for all detectors. At this time, in some detectors, the result for 60Co shows a 10% higher error, which is estimated to be due to the random summing. It was not considered in tuning the simulation geometry, but it was found that improvements were needed to reflect the coincidence summing when construction the virtual spectrum in the future. The determined simulation geometry was applied to generate theoretical gamma-ray energy spectra of representative artificial radionuclides. In order to create a virtual spectrum similar to the real one, the background spectrum was measured for each detector without a source, and the simulation results were calculated in the form of having the same energy channel as the background spectrum. The background spectrum and theoretical spectra of artificial radionuclides for each detector were databased so that virtual spectra could be generated under desired conditions. The virtual spectrum was generated by adding a background spectrum and a spectrum obtained by multiplying the spectrum of the desired nuclide by the concentration of the nuclide. The validity of generated virtual spectra was verified using the pre-developed gamma spectrometry program. As a results of gamma spectrometry of virtual spectra, the virtual spectra was verified by showing a difference within 20% from the radioactivity value input when generating the virtual spectra.