• Title/Summary/Keyword: Fast neutrons

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A Study on the Technology of Measuring and Analyzing Neutrons and Gamma-Rays Using a CZT Semiconductor Detector (CZT 반도체 검출기를 활용한 중성자 및 감마선 측정과 분석 기술에 관한 연구)

  • Jin, Dong-Sik;Hong, Yong-Ho;Kim, Hui-Gyeong;Kwak, Sang-Soo;Lee, Jae-Geun
    • 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.

Neutron Personal Dose Equivalent Evaluation Using Panasonic UD-809P Type TLD Albedo Dosimeters (Panasonic UD-809P 알비도 열형광선량계를 이용한 중성자 개인선량당량 평가)

  • Shin, Sang-Woon;Son, Joong-Kwon;Jin, Hua
    • 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.

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Analysis for the secondary gamma-ray emission for glasses irradiated with various doses of fast neutron: Case study borate and silicate glasses

  • O.L. Tashlykov;V. Yu. Litovchenko;N.M. Aristov;K.A. Mahmoud
    • Nuclear Engineering and Technology
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    • v.55 no.7
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    • pp.2366-2372
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    • 2023
  • Are borate and silicate glasses suitable for working as shieling materials against fast neutrons? To correctly answer the above question, some silicate, and borate-based glasses were fabricated and irradiated with various doses of fast neutrons varied between 1.73 and 12.10 MGy. The color and hardness of the fabricated glasses were affected by the fast neutron fluence where the transparent glasses turned colored as well as the hardness of the fabricated glasses was decreased. The gamma-ray spectrometric analysis shows a high activity concentration produced in the barium borate glasses due to the formation of radioisotopes Ba-131 and Ba-133 reaches to 5.92E+05 Bq and 4.25E+03 Bq, respectively for sample Cd-5 Batch 3. Additionally, the gamma-ray spectrometric analysis for the sodium silicate glasses shows low activity concentrations emitted from isotopes formed due to the activation of Y2O3-associated impurities. These activities are low compared to that emitted by barium borate-based glasses.

Neutron dosimetry with a pair of TLDs for the Elekta Precise medical linac and the evaluation of optimum moderator thickness for the conversion of fast to thermal neutrons

  • Marziyeh Behmadi;Sara Mohammadi;Mohammad Ehsan Ravari;Aghil Mohammadi;Mahdy Ebrahimi Loushab;Mohammad Taghi Bahreyni Toossi;Mitra Ghergherehchi
    • 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.

A Study on the Neutron Dose Distribution in Case of 10 MV X-rays Radiotherapy (10MV X선 방사선 치료 시 중성자 선량 분포에 관한 연구)

  • Park, Cheol-Soo;Lim, Cheong-Hwan;Jung, Hong-Ryang;Shin, Seong-Soo
    • Journal of radiological science and technology
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    • v.31 no.4
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    • pp.415-417
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    • 2008
  • This study is to measure the radiation dose of neutrons generated by the particle accelerator during X-ray (photon) treatment with a neutron detection method by using CR-39, and to research how the generation of neutrons may incur problems associated with radiation doses for patient treatment when using high energy photons for cancer treatment as a clinical application. The findings are summarized as follows : The results showed that average 0.35mSv was measured with exposure of 1Gy photon in case of fast neutron, 0.65mSv with exposure of 2Gy photon, 1.82mSv exposure of 5Gy, 0.26mSv with exposure of 1Gy photon in case of thermal neutron, 0.56mSv with exposure of 2Gy photon, and 1.23mSv with exposure of 5Gy of photon. By measuring the occurrence of neutron by using Wedge Filter, it has been confirmed that the occurrence of neutrons increased when using Wedge Filter. The results also showed that more neutrons were detected over the existing experiments when using an SRS Cone requiring high doses of radiation. Total 2.85mSv neutrons were found on the average with exposure of 5Gy photon in case of fast neutron and 1.37mSv neutrons were found on the average with exposure of 5Gy photon in case of thermal neutron. During the general treatment, about 1.6 times more neutrons over 5Gy photon were found in case of fast neutron and about 1.12 time more neutrons over 5Gy photon were found in case of thermal neutron.

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Digital n-γ Pulse Shape Discrimination in Organic Scintillators with a High-Speed Digitizer

  • Kim, Chanho;Yeom, Jung-Yeol;Kim, Geehyun
    • 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.

Effects of Fast Neutron Irradiation on Switching of Silicon Bipolar Junction Transistor

  • Sung Ho Ahn;Gwang Min Sun
    • Journal of Radiation Protection and Research
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    • v.48 no.3
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    • pp.124-130
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    • 2023
  • Background: When bipolar junction transistors (BJTs) are used as switches, their switching characteristics can be deteriorated because the recombination time of the minority carriers is long during turn-off transient. When BJTs operate as low frequency switches, the power dissipation in the on-state is large. However, when BJTs operate as high frequency switches, the power dissipation during switching transients increases rapidly. Materials and Methods: When silicon (Si) BJTs are irradiated by fast neutrons, defects occur in the Si bulk, shortening the lifetime of the minority carriers. Fast neutron irradiation mainly creates displacement damage in the Si bulk rather than a total ionization dose effect. Defects caused by fast neutron irradiation shorten the lifetime of minority carriers of BJTs. Furthermore, these defects change the switching characteristics of BJTs. Results and Discussion: In this study, experimental results on the switching characteristics of a pnp Si BJT before and after fast neutron irradiation are presented. The results show that the switching characteristics are improved by fast neutron irradiation, but power dissipation in the on-state is large when the fast neutrons are irradiated excessively. Conclusion: The switching characteristics of a pnp Si BJT were improved by fast neutron irradiation.

SPECTRUM WEIGHTED RESPONSES OF SEVERAL DETECTORS IN MIXED FIELDS OF FAST AND THERMAL NEUTRONS

  • Kim, Sang In;Chang, Insu;Kim, Bong Hwan;Kim, Jang Lyul;Lee, Jung Il
    • Nuclear Engineering and Technology
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    • v.46 no.2
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    • pp.273-280
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    • 2014
  • The spectrum weighted responses of various detectors were calculated to provide guidance on the proper selection and use of survey instruments on the basis of their energy response characteristics on the neutron fields. To yield the spectrum weighted response, the detector response functions of 17 neutron-measuring devices were numerically folded with each of the produced calibration neutron spectra through the in-house developed software 'K-SWR'. The detectors' response functions were taken from the IAEA Technical Reports Series No. 403 (TRS-403). The reference neutron fields of 21 kinds with 2 spectra groups with different proportions of thermal and fast neutrons have been produced using neutrons from the $^{241}Am$-Be sources held in a graphite pile, a bare $^{241}Am$-Be source, and a DT neutron generator. Fluence-average energy ($E_{ave}$) varied from 3.8 MeV to 16.9 MeV, and the ambient-dose-equivalent rate [$H^*(10)/h$] varied from 0.99 to 16.5 mSv/h.

Electrical characteristics and deep-level transient spectroscopy of a fast-neutron-irradiated 4H-SiC Schottky barrier diode

  • Junesic Park;Byung-Gun Park;Hani Baek;Gwang-Min Sun
    • Nuclear Engineering and Technology
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    • v.55 no.1
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    • pp.201-208
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    • 2023
  • The dependence of the electrical characteristics on the fast neutron fluence of an epitaxial 4H-SiC Schottky barrier diode (SBD) was investigated. The 30 MeV cyclotron was used for fast neutron irradiation. The neutron fluences evaluated through Monte Carlo simulation were in the 2.7 × 1011 to 1.45 × 1013 neutrons/cm2 range. Current-voltage and capacitance-voltage measurements were performed to characterize the samples by extracting the parameters of the irradiated SBDs. Neutron-induced defects in the epitaxial layer were identified and quantified using a deep-level transient spectroscopy measurement system developed at the Korea Atomic Energy Research Institute. As the neutron fluence increased from 2.7 × 1011 to 1.45 × 1013 neutrons/cm2, the concentration of the Z1/2 defects increased by approximately 20 times. The maximum defect concentration was estimated as 1.5 × 1014 cm-3 at a neutron fluence of 1.45 × 1013 neutrons/cm2.

Evaluation of Neutron Shielding Performance of Polyethylene Coated Boron Carbide-Incorporated Cement Paste using MCNP Simulation (MCNP 시뮬레이션을 통한 폴리에틸렌 코팅 탄화붕소 혼입 시멘트 페이스트의 중성자 차폐 성능 평가)

  • Park, Jae-Yeon;Jee, Hyeon-Seok;Bae, Sung-Chul
    • Proceedings of the Korean Institute of Building Construction Conference
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    • 2018.11a
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    • pp.114-115
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    • 2018
  • To develop an effective shielding material for spent fuel that emits fast neutrons is necessary. In this study, thermal neutron and fast neutron shielding performance of polyethylene coated boron carbide-incorporated cement paste was quantitatively analyzed by Monte Carlo N-Particle transport code (MCNP) simulations. As the results of the simulations, fast neutrons were effectively shielded through large quantity of hydrogen and boron elements in polyethylene and boron carbide.

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