• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.1024-1029
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• 2022

Composite scintillation screens on a base of Gd_1.2Y_1.8Ga_2.5Al_2.5O₁₂:Ce (GYAGG) scintillator have been evaluated for neutron detection. Besides the powdered scintillator, the composite includes ⁶LiF particles; both are merged with a binder and deposited onto the light-reflecting aluminum substrate. Results obtained demonstrates that screens are suitable for use with a silicon photomultiplier readout to create a prospective solution for a compact and low-cost thermal neutron sensor. Composite GYAGG/⁶LiF scintillation screen shows a pretty matched sensitivity and γ-background rejection with a widely used ZnS/⁶LiF screens however, possesses forty times faster response.

• Radiation Oncology Journal
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• v.22 no.1
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• pp.64-68
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• 2004

Purpose : To study the feasibility of verifying real-time 2-D dose distribution measurement system with the scintillation screen for the quality assurance. Materials and Methods : The water phantom consisted of a scintillation screen (LANEX fast screen, Kodak, USA) that was axially located in the middle of an acrylic cylinder with a diameter of 25 cm. The charge-coupled device (CCD) camera was attached to the phantom In order to capture the visible light from the scintillation screen. To observe the dose distribution In real time, the intensity of the light from the scintillator was converted to a dosage. The isodose contours of the calculations from RTP and those of the measurements using the scintillation screen were compared for the arc therapy and the Intensity modulated radiation therapy (IMRT). Results : The kernel, expressed as a multiplication of two error functions, was obtained in order to correct the sensitivity of the CCD of the camera and the scintillation screen. When comparing the calculated isodose and measured isodose, a discrepancy of less than 8 mm in the high dose region was observed. Conclusion : Using the 2-D dosimetry system, the relationship between the light and the dosage could be found, and real-time verification of the dose distribution was feasible.

• 한국정보디스플레이학회:학술대회논문집
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• 2006.08a
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• pp.1239-1242
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• 2006

Scintillation that is grainy patterns appeared on a screen has been one of a biggest issues in a rear projection TVs. In this paper, with focusing on the average size of random particle, it was proved that the particle size of calculated speckle and the one of measured scintillation are almost the equal. This result shows speckle phenomenon is an important factor of scintillation.

• Proceedings of the Korean Society of Medical Physics Conference
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• 2006.08a
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• pp.72-72
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• 2006

• Radiation Oncology Journal
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• v.17 no.3
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• pp.249-255
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• 1999

Purpose : The purpose of this study is to investigate fundamental aspects of the dose response of fluorescent screen-based electronic portal imaging devices (EPIDS). Materials and Methods : We acquired scanned signal across portal planes as we varied the radiation that entered the EPID by changing the thickness and anatomy of the phantom as well as the air gap between the phantom and the EPID. In addition, we simulated the relative contribution of the scintillation light signal in the EPID system. Results : We have shown that the dose profile across portal planes is a function of the air gap and phantom thickness. We have also found that depending on the density change within the phantom geometry, errors associated with dose response based on the EPID scan can be as high as $7\%$. We also found that scintillation light scattering within the EPID system is an important source of error. Conclusion : This study revealed and demonstrated fundamental characteristics of dose response of EPID, as relative to that of ion chambers. This study showed that EPID based on fluorescent screen cannot be an accurate dosimetry system.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3133-3139
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• 2023

Neutron imaging technology as a means of non-destructive detection of materials is complementary to X-ray imaging. Silicon photomultiplier (SiPM), a new type of optical readout device, has overcome some shortcomings of traditional photomultiplier tube (PMT), such as high-power consumption, large volume, high price, uneven gain response, and inability to work in strong magnetic fields. Its application in the field of neutron detection will be an irresistible general trend. In this paper, a thermal neutron imaging detector based on ⁶LiF/ZnS scintillation screen and SiPM array readout was developed. The design of the detector geometry was optimized by geant4 Monte Carlo simulation software. The optimized detector was evaluated with a step wedge sample. The results show that the detector prototype with a 48 mm × 48 mm sensitive area can achieve about 38% detection efficiency and 0.26 mm position resolution when using a 300 ㎛ thick ⁶LiF/ZnS scintillation screen and a 2 mm thick Bk7 optical guide coupled with SiPM array, and has good neutron imaging capability. It provides effective data support for developing high-performance imaging detectors applied to the China Spallation Neutron Source (CSNS).

• Applied Chemistry for Engineering
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• v.17 no.6
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• pp.604-608
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• 2006

Polymer composites for measuring the radioactive contamination are prepared by coating ZnS(Ag) powders as a scintillator on polysulfone base layer. The composites consist of the active layer for a scintillation reaction with radioactive wastes and the transparent support layer for transmittance of light photons emitted by scintillation in the active layer. The binding of the active layer, including ZnS(Ag), on the support layer is proceeded via coating with polysulfone as a binder, without any extra adhesive. The coating was obtained by either casting via a Doctor Blade as applicator or screen printing. The prepared composites feature a monolithic structure, resulting in the complete adhesion between two layers. The composite prepared by the casting technique using an applicator holds a good detection efficiency in measuring the alpha radionuclide, but its structure becomes fragile because of warping in morphology. On the contrary, the composite prepared by the screen printing shows a good detection capacity as well as a good stability in a mechanical shape.

• Nuclear Engineering and Technology
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• v.56 no.5
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• pp.1836-1844
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• 2024

The use of digital systems in radiation science has been increased last years in the different knowledge areas, as a detectors, spectrometry, spectroscopy, simulation, etc. This manuscript presents the design and implementation of a low-cost, fully portable multi-channel analyzer for nuclear spectrometry (in situ). The development is based on a 32-bit microcontroller with ARM Cortex-M7, this design is able to digitize and analyze pulses from a radiation detector without the need to transform the input signal with some filter, obtains the maximum height of each of the digitized pulses, segmenting the information into channels to form a histogram and visualizing the LCD screen incorporated in the system. A continuous digitization methodology was used, which is in charge of the DMA and an ADC with a resolution of 12 bits at a speed of 3.6 MSPS. The system has a compact design and can open and save spectra in an SD memory built into the system. The MCA in MCU was tested with a NaI(Tl) Scintillation radiation detector, which allowed us to determine that the spectra obtained are similar compared to commercial MCA's. The results obtained show that the MCA in MCU is efficient for nuclear spectrometry, in addition to being very economical and low power consumption.

• Journal of IKEEE
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• v.26 no.3
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• pp.510-514
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• 2022

In this paper, we propose the development of high-sensitivity low-end radiation measuring sensor module. The proposed measurement sensor module is a scintillator + photomultiplier(SiPM) sensor optimization structure design, amplification and filter and control circuit design for sensor driver, control circuit design including short-distance communication, sensor mechanism design and manufacturing, and GUI development applied to prototypes consists of, etc. The scintillator + photomultiplier(SiPM) sensor optimization structure design is designed by checking the characteristics of the scintillator and the photomultiplier (SiPM) for the sensor structure design. Amplification, filter and control circuit design for sensor driver is designed to process fine scintillation signal generated by radiation with a scintillator using SiPM. Control circuit design including short-distance communication is designed to enable data transmission through MCU design to support short-range wireless communication function and wired communication support. The sensor mechanism design and manufacture is designed so that the glare generated by wrapping a reflective paper (mirroring) on the outside of the plastic scintillator is reflected to increase the efficiency in order to transmit the fine scintillation signal generated from the plastic scintillator to the photomultiplier(SiPM). The GUI development applied to the prototype expresses the date and time at the top according to each screen and allows the measurement unit and time, seconds, alarm level, communication status, battery capacity, etc. to be expressed. In order to evaluate the performance of the proposed system, the results of experiments conducted by an authorized testing institute showed that the radiation dose measurement range was 30 𝜇Sv/h ~ 10 mSv/h, so the results are the same as the highest level among products sold commercially at domestic and foreign. In addition, it was confirmed that the measurement uncertainty of ±7.4% was measured, and normal operation was performed under the international standard ±15%.

• Journal of the Korean Society for Nondestructive Testing
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• v.14 no.4
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• pp.228-236
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• 1995

In this paper the method to evaluate shielding deficiency by gamma scanning test was presented and verified theoretically by Monte Carlo code which is one of the best effective method for radiation shielding calculation. The cylindrical shielding model was selected to evaluate shielding deficiency by gamma scanning test. First, the reference shielding according to the design requirement of cask was fabricated specially and reference values were measured with Co-60 source and scintillation detector. As a result with which calculated the reference values, it is shown that maximum deficiency thickness for lead of true cylindrical shielding model was 12mm. To verify this, thickness of lead was calculated by MCNP code and maximum deficiency thickness was 11.6mm. The experimental result obtained by the use of reference shielding was in good agreement with the theoretical result within 4.1%. So, this method can be applied to inspect the shielding ability for great shielding or cask which the radioactive material is used. To perform measurement more exactly, the further work on the development of measuring equipment to display the results on the screen will be required.