• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2586-2591
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• 2022

In the present work, we have irradiated the DI-BSCCO superconducting tapes with the 100 keV deuterium ions to investigate the effect of ion irradiation on their critical current (Ic). The damage simulations are carried out using the binary collision approximation method to get the spatial distribution and depth profile of the damage events in the high temperature superconducting (HTS) tape. The point defects are formed near the surface of the HTS tape. These point defects change the vortex profile in the superconducting tape. Due to the long-range interaction of vortices with each other, the Ic of the tape degrades at the 77 K and self magnetic field. The radiation dose of 2.90 MGy degrades the 44% critical current of the tape. The results of the displacement per atom (dpa) and dose deposited by the deuterium ions are used to fit an empirical relation for predicting the degradation of the Ic of the tape. We include the dpa, dose and columnar defect terms produced by the incident particles in the empirical relation. The fitted empirical relation predicts that light ion irradiation degrades the Ic in the DI-BSCCO tape at the self field. This empirical relation can also be used in neutron irradiation to predict the lifetime of the DI-BSCCO tape. The change in the Ic of the DI-BSCCO tape due to deuterium irradiation is compared with the other second-generation HTS tape irradiated with energetic radiation.

• Journal of the Korean Society of Radiology
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• v.17 no.3
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• pp.481-487
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• 2023

In this study, we analyzed the incidence of side effects of photoneutron dose according to the number of beams during intensity-modulated radiotherapy of prostate cancer using 15 MV. The radiation treatment plan design for intensity-modulated radiation therapy for prostate cancer was established with a prescription dose of 220 cGy per dose and a total of 7260 cGy for 33 treatments. The linear accelerator used in the experiment is Varian's True Beam STx (Varian, USA). Photoneutron dose was generated by using 15 MV energy in the planning target volume (PTV). The treatment plan was designed with IMRT 5, 7, and 9 portals using the Eclipse System (Varian Ver 10.0, USA). An optically stimulated luminescence albedo neutron dosimeter (Landauer Inc., USA) was used to measure photoneutron dose. IMRT 5 portals, 1.7 per 1,000, 7 portals, 1.8 per 1,000, 9 portals, 2.0 per 1,000 were calculated as the probability of experiencing side effects on the thyroid gland due to photoneutron dose. This study studies the risk of secondary radiation exposure dose that can occur during intensity-modulated radiation therapy, and it is considered that it will be used as useful data in relation to stochastic effects in the future.

• Journal of IKEEE
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• v.21 no.4
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• pp.408-411
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• 2017

In this paper, we propose an integrated control system that measures neutrons, gamma ray, and x-ray. The proposed system is able to monitor and control the data measured and analyzed on the remote or network, and can monitor and control the status of each part of the system remotely without remote control. The proposed system consists of a gamma ray/x-ray sensor part, a neutron sensor part, a main control embedded system part, a dedicated display device and GUI part, and a remote UI part. The gamma ray/x-ray sensor part measures gamma ray and x-ray of low level by using NaI(Tl) scintillation detector. The neutron sensor part measures neutrons using Proportional Counter Detector(low-level neutron) and Ion Chamber Type Detector(high-level neutron). The main control embedded system part detects radiation, samples it in seconds, and converts it into radiation dose for accumulated pulse and current values. The dedicated display device and the GUI part output the radiation measurement result and the converted radiation amount and radiation amount measurement value and provide the user with the control condition setting and the calibration function for the detection part. The remote UI unit collects and stores the measured values and transmits them to the remote monitoring system. In order to evaluate the performance of the proposed system, the measurement uncertainty of the neutron detector was measured to less than ${\pm}8.2%$ and the gamma ray and x-ray detector had the uncertainty of less than 7.5%. It was confirmed that the normal operation was not less than ${\pm}15$ percent of the international standard.

• 제어로봇시스템학회:학술대회논문집
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• 2003.10a
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• pp.1220-1224
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• 2003

In this paper, we measured the radiation dose of a fuel handling machine of the CANDU type Wolsong nuclear reactor directly during operation, in spite of the high radiation level. In this paper we will describe the sensor development, measurement techniques, and results of our study. For this study, we used specially developed semiconductor sensors and matching dosimetry techniques for the mixed radiation field. MOSFET dosimeters with a thin oxide, that are tuned to a high dose, were used to measure the ionizing radiation dose. Silicon diode dosimeters with an optimum area to thickness ratio were used for the radiation damage measurements. The sensors are able to distinguish neutrons from gamma/X-rays. To measure the radiation dose, electronic sensor modules were installed on two locations of the fuel handling machine. The measurements were performed throughout one reactor maintenance cycle. The resultant annual cumulative dose of gamma/X-rays on the two spots of the fuel handling machine were 18.47 Mrad and 76.50 Mrad, and those of the neutrons were 17.51 krad and 60.67 krad. The measured radiation level is high enough to degrade certain cable insulation materials that may result in electrical insulation failure.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.1
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• pp.133-140
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• 2021

This study evaluates the radioactivity of concrete waste that occurs due to large amounts of decommissioned nuclear wastes and then determines the surface dose rate when the waste is packaged in a disposal container. The radiation assessment was conducted under the presumption that impurities included in the bio-shielded concrete contain the highest amount of radioactivity among all the concrete wastes. Neutron flux was applied using the simplified model approach in a sample containing the most Co and Eu impurities, and a maximum of 9.8×104 Bq·g-1 60Co and 2.63×105 Bq·g-1 152Eu was determined. Subsequently, the surface dose rate of the container was measured assuming that the bio-shield concrete waste would be packaged in a newly developed disposal container. Results showed that most of the concrete wastes with a depth of 20 cm or higher from the concrete surface was found to have less than 1.8 mSv·hr-1 in the surface dose of the new-type disposal container. Hence, when bio-shielded concrete wastes, having the highest radioactivity, is disposed in the new disposal container, it satisfies the limit of the surface dose rate (i.e., 2 mSv·hr-1) as per global standards.

• Nuclear Engineering and Technology
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• v.51 no.8
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• pp.1964-1969
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• 2019

The safety and efficiency of nuclear reactors largely depend on the monitoring and control of nuclear radiation. Due to the unique nuclear-physical characteristics, Hf is one of the most promising materials for the manufacturing of the control rods and the emitters of neutron detectors. It is proposed to use the Compton neutron detector with the emitter made of Hf in the In-core Instrumentation System (ICIS) for monitoring the neutron field. The main advantages of such a detector in comparison the conventional β-emission sensors are the possibility of reaching of a higher cumulative radiation dose and the absence of signal delays. The response time of the detection is extremely important when a nuclear reactor is operating near its critical operational parameters. Taking Hf as an example, the general principles for calculating the chains of materials transformation under neutron irradiation are reported. The influence of ^179m1Hf on the Hf composition changing dynamics and the process of transmutants' (Ta, W) generation were determined. The effect of these processes on the absorbing properties of Hf, which inevitably predetermine the lifetime of the detector and its ability to generate a signal, is estimated.

• Nuclear Engineering and Technology
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• v.46 no.4
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• pp.547-556
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• 2014

Korea expects a shortage in storage capacity for spent fuels at reactor sites. Therefore, a need for more metal and/or concrete casks for storage systems is anticipated for either the reactor site or away from the reactor for interim storage. For the purpose of interim storage and transportation, a dual purpose metal cask that can load 21 spent fuel assemblies is being developed by Korea Radioactive Waste Management Corporation (KRMC) in Korea. At first the gamma and neutron flux for the design basis fuel were determined assuming in-core environment (the temperature, pressure, etc. of the moderator, boron, cladding, $UO_2$ pellets) in which the design basis fuel is loaded, as input data. The evaluation simulated burnup up to 45,000 MWD/MTU and decay during ten years of cooling using the SAS2H/OGIGEN-S module of the SCALE5.1 system. The results from the source term evaluation were used as input data for the final shielding evaluation utilizing the MCNP Code, which yielded the effective dose rate. The design of the cask is based on the safety requirements for normal storage conditions under 10 CFR Part 72. A radiation shielding analysis of the metal storage cask optimized for loading 21 design basis fuels was performed for two cases; one for a single cask and the other for a $2{\times}10$ cask array. For the single cask, dose rates at the external surface of the metal cask, 1m and 2m away from the cask surface, were evaluated. For the $2{\times}10$ cask array, dose rates at the center point of the array and at the center of the casks' height were evaluated. The results of the shielding analysis for the single cask show that dose rates were considerably higher at the lower side (from the bottom of the cask to the bottom of the neutron shielding) of the cask, at over 2mSv/hr at the external surface of the cask. However, this is not considered to be a significant issue since additional shielding will be installed at the storage facility. The shielding analysis results for the $2{\times}10$ cask array showed exponential decrease with distance off the sources. The controlled area boundary was calculated to be approximately 280m from the array, with a dose rate of 25mrem/yr. Actual dose rates within the controlled area boundary will be lower than 25mrem/yr, due to the decay of radioactivity of spent fuel in storage.

• Journal of the Korean Society of Radiology
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• v.16 no.4
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• pp.471-476
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• 2022

The purpose of this study was to study the probability of cancer occurrence due to photoneutron dose exposure of the colon and thyroid gland, which are normal organs, in 3D CRT, IMRT 5 portals, and IMRT 9 portals, which are radiotherapy methods for prostate cancer. The total prescribed dose for prostate cancer was 6600 cGy, 220 cGy per dose, and 30 divided irradiations were applied for the total number of times. After setting up the Rando phantom on the treatment table (couch) of the medical linear accelerator used in the experiment, an optically stimulated luminescence albedo neutron dosimeter was placed on the corresponding area of the large intestine and thyroid gland of the phantom for measurement. During 3D CRT of prostate cancer, the probability of secondary cancer due to photoneutron dose to the colon and thyroid gland, which are normal organs, was 1.8 per 10,000 people. And IMRT 5 portals were 8.7 per 10,000 people, which was about 5 times larger than 3D CRT. IMRT 9 portals derived the result that there is a probability that 1.2 people per 1,000 people will develop cancer. Based on this study, the risk of secondary radiation exposure due to the dose of photoneutrons generated during radiation therapy is studied, and it is thought that it will be used as useful data for radiation protection in relation to the stochastic effect of radiation in the future.

• Radiation Oncology Journal
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• v.30 no.1
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• pp.27-35
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• 2012

Purpose: To evaluate the effect of common three photon energies (6-MV, 10-MV, and 15-MV) on intensity-modulated radiation therapy (IMRT) plans to treat prostate cancer patients. Materials and Methods: Twenty patients with prostate cancer treated locally to 81.0 Gy were retrospectively studied. 6-MV, 10-MV, and 15-MV IMRT plans for each patient were generated using suitable planning objectives, dose constraints, and 8-field setting. The plans were analyzed in terms of dose-volume histogram for the target coverage, dose conformity, organs at risk (OAR) sparing, and normal tissue integral dose. Results: Regardless of the energies chosen at the plans, the target coverage, conformity, and homogeneity of the plans were similar. However, there was a significant dose increase in rectal wall and femoral heads for 6-MV compared to those for 10-MV and 15-MV. The $V_{20Gy}$ of rectal wall with 6-MV, 10-MV, and 15-MV were 95.6%, 88.4%, and 89.4% while the mean dose to femoral heads were 31.7, 25.9, and 26.3 Gy, respectively. Integral doses to the normal tissues in higher energy (10-MV and 15-MV) plans were reduced by about 7%. Overall, integral doses in mid and low dose regions in 6-MV plans were increased by up to 13%. Conclusion: In this study, 10-MV prostate IMRT plans showed better OAR sparing and less integral doses than the 6-MV. The biological and clinical significance of this finding remains to be determined afterward, considering neutron dose contribution.

• Progress in Medical Physics
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• v.15 no.4
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• pp.210-214
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• 2004

Since production of radioactive isotope for using PET, a lot of neutrons were produced. The produced neutrons were mainly shielded by concrete facility. Secondary photons are generated and emitted from the concrete shielding wall of the PET cyclotron since the proton-generated neutrons are thermalized and absorbed in the concrete wall and emit secondary radiations, i.e., photons. This study calculated neutron dose and photon dose at outside of the accelerator facility using MCNPX code. As results of the calculation, total dose were calculated less than limited dose by law.