• Journal of Radiation Protection and Research
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• v.24 no.3
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• pp.143-154
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• 1999

Panasonic UD-809P type albedo neutron TL dosimeters mounted on a water phantom were used to measure neutron personal dose equivalent in a Korean nuclear power plant. From the measured TL readings, personal dose equivalents from thermal, epithermal and fast neutrons were evaluated by using a method adopted in a neutron dose calculation algorithm for Panasonic UD-809P type albedo neutron TL dosimeters, which was suggested in a Panasonic TLD System User's Manual. The results showed that personal dose equivalent from fast neutrons could not be adequately evaluated in a field with high thermal neutron fraction to be encountered in a nuclear power plant. This seems to be related to the incomplete incidence of albedo thermal neutrons to the TL dosimeters. In order to evaluate appropriately the personal dose equivalent from fast neutrons in the field condition, new method fer the neutron dose calculation algorithm was suggested. In this new method, neutrons are grouped into thermal neutrons and fast neutrons. For each neutron component, equations for TL response, sensitivity factor, calibration factor and personal dose equivalent were derived.

• Journal of radiological science and technology
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• v.45 no.1
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• pp.57-67
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• 2022

CZT detectors, which are compound semiconductors that have been widely used recently for gamma-ray detection purposes, are difficult to detect neutrons because direct interaction with them does not occur unlike gamma-rays. In this paper, a method of detecting and determining energy levels (fast neutrons and thermal neutrons) of neutrons, in addition of identifying energy and nuclide of gamma-rays, and evaluating gamma dose rates using a CZT semiconductor detector is described. Neutrons may be detected by a secondary photoelectric effect or compton scattering process with a characteristic gamma-ray of 558.6 keV generated by a capture reaction (¹¹³Cd + ¹n → ¹¹⁴Cd + 𝛾) with cadmium (Cd) in the CZT detector. However, in the case of fast neutrons, the probability of capture reaction with cadmium (Cd) is very low, so it must be moderated to thermal neutrons using a moderator and the material and thickness of moderator should be determined in consideration of the portability and detection efficiency of the equipment. Conversely, in the case of thermal neutrons, the detection efficiency decreases due to shielding effect of moderator itself, so additional CZT detector that do not contain moderator must be configured. The CZT detector that does not contain moderator can be used to evaluate energy, nuclide, and gamma dose-rate for gamma-rays. The technology proposed in this paper provides a method for detecting both neutrons and gamma-rays using a CZT detector.

• Journal of Radiation Protection and Research
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• v.40 no.3
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• pp.147-154
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• 2015

The scatter photons and photoneutrons from high energy photon beams (more than 10 MV) will increase the undesired dose to the patient and the staff working in linear accelerator room. This undesired dose which is found at out-of-field area can increase the probability of secondary malignancy. The purpose of this study is to determine the equivalent dose of scatter photons and neutrons generated by 3 different treatment techniques: 3D-conformal, intensity modulated radiation therapy (IMRT) and volumetric modulated arc therapy (VMAT). The measurement was performed using two types of the optically stimulation luminescence detectors (OSL and OSLN) in the Alderson Rando phantom that was irradiated by 3 different treatment techniques following the actual prostate cancer treatment plans. The scatter photon and neutron equivalent dose were compared among the 3 treatments techniques at the surface in the out-of-field area and the critical organs. Maximum equivalent dose of scatter photons and neutrons was found when using the IMRT technique. The scatter neutrons showed average equivalent doses of 0.26, 0.63 and $0.31mSv{\cdot}Gy^{-1}$ at abdominal surface region which was 20 cm from isocenter for 3D, IMRT and VMAT, respectively. The scattered photons equivalent doses were 6.94, 10.17 and $6.56mSv{\cdot}Gy^{-1}$ for 3D, IMRT and VMAT, respectively. For the 5 organ dose measurements, the scattered neutron and photon equivalent doses in out of field from the IMRT plan were highest. The result revealed that the scatter equivalent doses for neutron and photon were higher for IMRT. So the suitable treatment techniques should be selected to benefit the patient and the treatment room staff.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3704-3710
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• 2021

The characteristics of the external radiation on the surface of the casks for spent nuclear fuel (SNF) storage by dry method are investigated for the case when the spatial distribution of SNF in the basket changes due to the destruction of the fuel rod claddings. The surface areas are determined, where the changes in fluxes of neutrons, produced by ²⁴⁴Cm actinide, and γ-quanta, produced by long-lived isotopes, are maximum in the result of the decrease in the height of the SNF area. Concrete (VSC-24) and metal (SC-21) casks are considered as examples. The procedure of periodic measurement of the dose rate of neutrons or γ-quanta at the specified points of the cask surface is proposed for identifying the fuel rod cladding destruction. Under normal operation, the decrease in the dose rate produced by neutrons as the function of SNF storage duration is determined by the half-life of ²⁴⁴Cm, and for γ-quanta - by the half-lives of long-lived SNF isotopes. Consequently, a stepwise change in the dose rate of neutrons or γ-quanta, detected by the measurements, as compared to the previous one, would indicate the destruction of the fuel rod claddings.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.753-761
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• 2024

Introduction: In this study, TLD 600 and TLD 700 pairs were used to measure the neutron dose of Elekta Precise medical linac. To this end, the optimum moderate thickness for the conversion of fast to thermal neutrons were evaluated. Materials and methods: ²⁴¹Am-Be and ²⁵²Cf sources were simulated to calculate the optimum thicknesses of the moderator for the conversion of maximum fast neutrons (FN) into thermal neutrons (TN). Pair TLDs were used to measure F&TN doses for three different field sizes at four depths of the medical linac. Results: The maximum thickness of the moderator was optimized at 6 cm. The measurement results demonstrated that the TN dose increased with the expansion of field size and depth. The FN dose, which was converted TN, exhibits behaviors comparable to the TN due to its nature. Conclusion: This study presents the optimum thickness for the moderator to convert FN into TN and measure F&TN using TLDs.

• Journal of Radiation Protection and Research
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• v.23 no.1
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• pp.17-23
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• 1998

Dose response relationships for one of biological end-points (gene mutation) in somatic cells of Tradescantia 4430 clones were studied using neutrons coming out of a californium-252 isotopic source. And the relative biological effectiveness (RBE) of neutrons in relation to X-rays in the induction of TSH pink mutations was assessed. Inflorescences were irradiated with X-ray from X-ray generator and neutrons from $^{252}Cf$ source. Irradiated cuttings were incubated with aeration in neutrient solution under the controlled condition. For more than 4 weeks after irradiation cell mutations were scored. Pink mutation frequencies were calculated from the pooled data for the peak interval (days 6 to 13 post-irradiation). Somatic cell mutations in TSH showed linear dose response relationships in the range of neutron doses available for the experiment. The RBE values estimated for neutrons in relation to X-rays were in the range 3.1 to 6.8, which were much lower than normally recognized value.

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1425-1430
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• 2024

This study develops superheated emulsion detectors that are both sensitive to fast neutrons, and thermal neutrons owing to the exergonic ⁶₃Li(n, α)³₁H capture reaction caused by the ⁶Li-containing compound dispersed throughout the gel-like medium. The experimental research was conducted on two SEDs. One detector was an ordinary Freon-12 detector and the other was a Freon-12 detector containing 3.4 % (by weight) LiCl. In order to investigate the sensitivity of lithium-containing SEDs to thermal neutrons, two types of SEDs were simultaneously exposed to various flux levels of thermal neutrons from ²⁴¹Am-Be neutron source inside a cylindrical tank filled with water. A Boron-lined proportional counter was used to estimate the thermal neutron flux and the relevant MCNP code was developed for flux and dose calculations in the prepared set-up around ²⁴¹Am-Be source. The results demonstrate that there is a proportional relationship between the variations of SED response and the change in thermal neutron flux and dose. Also, the sensitivity of SED was estimated.

• 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.

• Proceedings of the Korean Society of Medical Physics Conference
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• 2002.09a
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• pp.157-160
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• 2002

The dosage of intra-operative BNCT using near-threshold $^{7}$ Li(p,n)$^{7}$ Be direct neutrons was evaluated with the calculation method validated with the phantom experiment. The production of both neutrons by near-threshold $^{7}$ Li(p,n)$^{7}$ Be and gamma rays by $^{7}$ Li(p,p'gamma)$^{7}$ Li in a Li target was calculated using Lee's method and their transport in the phantom was calculated with MCNP-4B. As a result, the region satisfying the requirements of the protocol in intra-operative BNCT for brain tumors in Japan was acknowledged to be comparable to present BNCT, for the proton energy of 1.900 MeV for example. A boron-dose enhancer (BDE) introduced in this study to increase $^{10}$ (n,$\alpha$)$^{7}$ Li dose in a living body was effective. The void used to increase doses in deep regions was also valid with the BDE. It was found that intra-operative BNCT using near-threshold $^{7}$ Li(p,n)$^{7}$ Be direct neutrons is feasible.

• Nuclear Engineering and Technology
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• v.50 no.6
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• pp.989-995
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• 2018

Argonne National Laboratory of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have cooperated on the development, design, and construction of a neutron source facility. The facility was constructed at Kharkov, Ukraine, and its commissioning process is underway. The facility will be used for researches, producing medical isotopes, and training young nuclear specialists. The neutron source facility is designed with a provision to include a cryogenically cooled moderator system-a cold neutron source (CNS). This CNS provides low-energy neutrons, which will be used in the scattering experiment and material structures analysis. Cold neutron guides, coated with reflective material for the low-energy neutrons, will be used to transport the cold neutrons to the experimental site. The cold neutron guides would keep the cold neutrons within certain energy and angular space concentrated inside, while most of the gamma rays and high-energy neutrons are not affected by the cold neutron guides. For the KIPT design, the cold neutron guides need to extend several meters outside the main shield of the facility, and curved guides will also be used to remove the gamma and high-energy neutron. The neutron guides should be installed inside a shield structure to ensure an acceptable biological dose in the facility hall. Heavy concrete is the selected shielding material because of its acceptable performance and cost. Shield design analysis was carried out for the CNS guide hall. MCNPX was used as the major computation tool for the design analysis, with neutron and gamma dose calculated separately. Weight windows variance reduction technique was also used in the shield design. The goal of the shield design is to keep the total radiation dose below the $5.0{\mu}Sv/hr$ guideline outside the shield boundary. After a series of iterative MCNPX calculations, the shield configuration and parameters of CNS guide hall were determined and presented in this article.