• Journal of Radiation Protection and Research
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• v.48 no.4
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• pp.184-196
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• 2023

Background: Radiation protection is crucial in various fields due to the harmful effects of radiation. Shielding is used to reduce radiation exposure, but gamma radiation poses challenges due to its high energy and penetration capabilities. Materials and Methods: This work investigates the radiation shielding properties of polyvinylidene fluoride (PVDF) samples containing different weight fraction of tungsten carbide (WC), tungsten trioxide (WO₃), and tungsten disulfide (WS₂). Parameters such as the mass attenuation coefficient (MAC), half-value layer (HVL), mean free path (MFP), effective atomic number (Z_eff), and macroscopic effective removal cross-section for fast neutrons (Σ_R) were calculated using the Phy-X/PSD software. EpiXS simulations were conducted for MAC validation. Results and Discussion: Increasing the weight fraction of the additives resulted in higher MAC values, indicating improved radiation shielding. PVDF-xWC showed the highest percentage increase in MAC values. MFP results indicated that PVDF-0.20WC has the lowest values, suggesting superior shielding properties compared to PVDF-0.20WO₃ and PVDF-0.20WS₂. PVDF-0.20WC also exhibited the highest Z_eff values, while PVDF-0.20WS₂ showed a slightly higher increase in Z_eff at energies of 0.662 and 1.333 MeV. PVDF-0.20WC has demonstrated the highest Σ_R value, indicating effective shielding against fast neutrons, while PVDF-0.20WS₂ had the lowest Σ_R value. The Monte Carlo N-Particle Transport version 5 (MCNP5) simulations showed that PVDF-xWC attenuates gamma radiation more than pure PVDF, significantly decreasing the dose equivalent rate. Conclusion: Overall, this research provides insights into the radiation shielding properties of PVDF mixtures, with PVDF-xWC showing the most promising results.

• Nuclear Engineering and Technology
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• v.12 no.3
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• pp.171-179
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• 1980

This paper presentss flux-to-dose conversion factors for neutrons and gamma-rays based on the concept of the maximum absorbed dose. Neutron flux-to-does-rate conversion factors for energies from 2.5$\times$10$^{-8}$ to 20 MeV are presented while the conversion factors for gamma-rays are given in the energy range of 0.01 to 15MeV. Flux-to-does-rate conversion factors, which were calculated under the assumption that the radiation energy distribution has nonlinearity in phantom, are different from those values obtained by monoenergetic radiation. Especially, these values obtained here were determined for the cross section libray such as DLC-23, DLC-27, and DLC-31. The flux-to-dose-rate conversion factors obtained in this work are in a good agreement with the values presented by American National Standard Institute (ANSI) N666. These results are used to calculate the dose rate distribution of neutron and gamma-ray in any radiation fields, and will be useful for the radiation shielding analysis, radiation protection and radiation dosimetry concerned with problems of continuous energy distribution.

• Journal of Radiation Protection and Research
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• v.26 no.3
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• pp.203-206
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• 2001

The mathematical processing (unfolding) of pulse height spectra from a scintillation detector helps to calculate the photon fluence rate energy distribution in a measured photon field. The data processing is based on the knowledge of detection system response function and directional dependence respectively. The experimental results of the photon fields measurements in the vicinity of the spent fuel temporary storage and inside the storage hall are presented. The containers Castor 440 are used for temporary storing of the burnt up fuel assemblies in the Czech nuclear power plant Dukovany. A set of periodical measurements was performed in order to get basic information on the time dependence of the photon fields spatial distributions and spectral characteristics in the temporary storage hall and its vicinity. The photon fields were measured by the scintillation system. The obtained photon fields spatial distributions and spectral characteristics present the information on the radiation hazard in the storage.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.1045-1051
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• 2023

In this study, we fabricated a liquid light-guide-based radiation sensor with a LaBr₃:Ce scintillator for remote gamma-ray spectroscopy. We acquired the energy spectra of Cs-137 and Co-60 using the proposed sensor, estimated the energy resolutions of the full energy peaks, and compared the scintillation light output variations. The major peaks of the radionuclides were observed in each result, and the estimated energy resolutions were similar to that of a general NaI(Tl) scintillation detector without a liquid light guide. Moreover, we showed the relationships of energy resolution and analog-to-digital channel regarding the number of photoelectrons produced and confirmed the effects of light guide length on remote gamma-ray spectroscopy. The proposed sensor is expected to be utilized to perform remote gamma-ray spectroscopy for distances of 3 m or more and would find application in many fields of nuclear facilities and industry.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.283-292
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• 2022

Advanced radiation applications have been widely used and extended to many fields. As a result of this fact, choosing an appropriate shielding material based on the radiation application has become vital. In this regard, the integration of elements into polymer composites has been investigated and contributed to the quantity and quality of radiation shielding materials. This study reports photon attenuation parameters and electromagnetic shielding effectiveness of a novel polymer composite prepared with a matrix reinforced with three different proportions (5, 10, and 15 wt%) of niobium content. Addition of Nb dopant improves both photon attenuation and electromagnetic shielding effectiveness for the investigated composites. Therefore, Nb(15%) polymer composite with highest concentration has been found to be the best absorber for ionizing and non-ionizing radiations. Consequently, the performed analyzes provide evidences that the prepared Nb-reinforced polymer composite could be effectively used as photon radiation attenuator and electromagnetic shielding material.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.685-687
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• 2024

Fiber optic cables are used extensively for remote monitoring in applications under extreme conditions, such as at high temperatures or in ionizing radiation fields. When high temperature fiber optic cables were subjected to gamma irradiations, there was a significant loss in transmission at wavelengths < 350 nm after only 1 minute of irradiation. Negligible recovery of the fiber optic transmission with time was observed over 2 years, but the irradiation damage was almost completely reversed by high temperature annealing at 400 ℃.

• Radiation Oncology Journal
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• v.27 no.2
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• pp.91-102
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• 2009

Purpose: To compare the accuracy and efficacy of EDR2 film, a 2D ionization chamber array (MatriXX) and an amorphous silicon electronic portal imaging device (EPID) in the pre-treatment QA of IMRT. Materials and Methods: Fluence patterns, shaped as a wedge with 10 steps (segments) by a multi-leaf collimator (MLC), of reference and test IMRT fields were measured using EDR2 film, the MatriXX, and EPID. Test fields were designed to simulate leaf positioning errors. The absolute dose at a point in each step of the reference fields was measured in a water phantom with an ionization chamber and was compared to the dose obtained with the use of EDR2 film, the MatriXX and EPID. For qualitative analysis, all measured fluence patterns of both reference and test fields were compared with calculated dose maps from a radiation treatment planning system (Pinnacle, Philips, USA) using profiles and $\gamma$ evaluation with 3%/3 mm and 2%/2 mm criteria. By measurement of the time to perform QA, we compared the workload of EDR2 film, the MatriXX and EPID. Results: The percent absolute dose difference between the measured and ionization chamber dose was within 1% for the EPID, 2% for the MatriXX and 3% for EDR2 film. The percentage of pixels with $\gamma$%>1 for the 3%/3 mm and 2%/2 mm criteria was within 2% for use of both EDR2 film and the EPID. However, differences for the use of the MatriXX were seen with a maximum difference as great as 5.94% with the 2%/2 mm criteria. For the test fields, EDR2 film and EPID could detect leaf-positioning errors on the order of -3 mm and -2 mm, respectively. However it was difficult to differentiate leaf-positioning errors with the MatriXX due to its poor resolution. The approximate time to perform QA was 110 minutes for the use of EDR2 film, 80 minutes for the use of the MatriXX and approximately 55 minutes for the use of the EPID. Conclusion: This study has evaluated the accuracy and efficacy of EDR2 film, the MatriXX and EPID in the pre-treatment verification of IMRT. EDR2 film and the EPID showed better performance for accuracy, while the use of the MatriXX significantly reduced measurement and analysis times. We propose practical and useful methods to establish an effective QA system in a clinical environment.

• Radiation Oncology Journal
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• v.2 no.1
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• pp.149-153
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• 1984

Postoperative radiotherapy of breast cancer makes it possible to reduce loco-regional recurrence of breast cancer. The treatment technique, which can reduce the low-dose region at the junction and lung, is required. To produce proper dose distribution of internal mammary chain and chest wall, authors tried to find the method to expose $^{60}Co\;\gamma-ray$ on internal mammary region and 7MeV electron on chest wall. Exposure time of $^{60}Co\;\gamma$ and monitor unit of 9MeV were selected so that dose of $^{60}Co$ at 4cm depth was the same as that of 7Mev electron at $80\%$ dose depth. The position and direction of electron beam were changed for $^{60}Co$ beam: $0^{\circ},\;5^{\circ}$ for 0cm seperation; $0^{\circ},\;5^{\circ},\;10^{\circ}$ for 0.5cm seperation; $5^{\circ},\;10^{\circ},\;15^{\circ}$ for 1cm seperation. The results are as followings. 1. When the seperation of two fields was increased, dose on the axis of $^{60}Co$ beam was increased and dose at the junction region decreased while the volume of lung to be exposed to high dose and hot spot size were irregularly changed. 2. The dose distribution in the target volume of internal mammary and chest wall was most ideal when the seperation of two fields was $0\~0.5cm$ and the direction of electron beam was parallel to $^{60}Co$ beam.

• Journal of Radiation Protection and Research
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• v.44 no.2
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• pp.53-63
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• 2019

Background: As neutron fields are always accompanied by gamma rays, it is essential to distinguish neutrons from gamma rays in the detection of neutrons. Neutrons and gamma rays can be separated by pulse shape discrimination (PSD) methods. Recently, we performed characterization of a stilbene scintillator detector and an EJ-301 liquid scintillator detector with a high-speed digitizer DT5730 and investigated optimized PSD variables for both detectors. This study is for providing a basis for developing fast neutron/gamma-ray dual-particle imager. Materials and Methods: We conducted PSD experiments using stilbene scintillator and EJ-301 liquid scintillator and evaluated neutron and gamma ray discriminability of each PSD method with a $^{137}Cs$ gamma source and a $^{252}Cf$ neutron source. We implemented digital signal processing techniques to apply two PSD methods - the charge comparison (CC) method and the constant time discrimination (CTD) method - to distinguish neutrons from gamma rays. We tried to find optimized PSD variables giving the best discriminability in a given experimental condition. Results and Discussion: For the stilbene scintillator detector, the charge comparison method and the constant time discrimination method both delivered the PSD FOM values of 1.7. For the EJ-301 liquid scintillator detector, both PSD methods delivered the PSD FOM values of 1.79. With the same PSD variables, PSD performance was excellent in $300{\pm}100keVee$, $500{\pm}100keVee$, and $700{\pm}100keVee$ energy regions. This result shows that we can achieve an effective discrimination of neutrons from gamma rays using these scintillator detector systems. Conclusion: We applied both PSD methods to a stilbene and a liquid scintillator and optimized the PSD performance represented by FOM values. We observed a good separation performance of both scintillators combined with a high-speed digitizer and digital PSD. These results will provide reference values for the dual-particle imager we are developing, which can image both fast neutrons and gamma rays simultaneously.

• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1421-1430
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• 2022

The computation of the solid angle and the detector efficiency is considering to be one of the most important factors during the measuring process for the radioactivity, especially the cylindrical γ-ray NaI(Tl) detectors nowadays have applications in several fields such as industry, hazardous for health, the gamma-ray radiation detectors grow to be the main essential instruments in radiation protection sector. In the present work, a generic numerical simulation method (NSM) for calculating the efficiency of the γ-ray spectrometry setup is established. The formulas are suitable for any type of source-to-detector shape and can be valuable to determine the full-energy peak and the total efficiencies and P/T ratio of cylindrical γ-ray NaI(Tl) detector setup concerning the truncated conical radioactive source. This methodology is based on estimate the path length of γ-ray radiation inside the detector active medium, inside the source itself, and the self-attenuation correction factors, which typically use to correct the sample attenuation of the original geometry source. The calculations can be completed in general by using extra reasonable and complicate analytical and numerical techniques than the standard models; especially the effective solid angle, and the detector efficiency have to be calculated in case of the truncated conical radioactive source studied condition. Moreover, the (NSM) can be used for the straight calculations of the γ-ray detector efficiency after the computation of improvement that need in the case of γ-γ coincidence summing (CS). The (NSM) confirmation of the development created by the efficiency transfer method has been achieved by comparing the results of the measuring truncated conical radioactive source with certified nuclide activities with the γ-ray NaI(Tl) detector, and a good agreement was obtained after corrections of (CS). The methodology can be unlimited to find the theoretical efficiencies and modifications equivalent to any geometry by essential sufficiently the physical selective considered situation.