• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1964-1969
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• 2021

We developed a Dual-PPACs detector for fast neutron measurements that consists of two sets of PPAC: conventional PPAC and fission PPAC. A²³⁸U(U₃O₈) coating is placed in the fission PPAC's anode, which is used as the neutrons conversion layer. An experiment was performed to measure neutron time-of-flight (TOF) in which ²⁵²Cf spontaneous fission source was used. An excellent time resolution of 164ps has been observed at 6 mbar in isobutene gas. With the excellent time resolution of Dual-PPACs detector, exact neutron energy can be extracted from the timing measurement. The experimental detection efficiency was 1.9 × 10^-7, consistent with the efficiency of 2.5 × 10^-7 given by a Geant4 simulation. Ultimately, the results show that the Dual-PPACs detector is a suitable candidate for measuring fast neutrons in the future CiADS system.

• 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.

• Progress in Medical Physics
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• v.27 no.3
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• pp.162-168
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• 2016

Proton therapy is increasingly being actively used in the treatment of cancer. In contrast to photons, protons have the potential advantage of delivering higher doses to the cancerous tissue and lower doses to the surrounding normal tissue. However, a range shifter is needed to degrade the beam energy in order to apply the pencil beam scanning technique to tumors located close to the minimum range. The secondary neutrons are produced in the beam path including within the patient's body as a result of nuclear interactions. Therefore, unintended side effects may possibly occur. The research related to the secondary neutrons generated during proton therapy has been presented in a variety of studies worldwide, since 2007. In this study, we measured the magnitude of the secondary neutron dose depending on the location of the detector and the use of a range shifter at the beam nozzle of the proton scanning mode, which was recently installed. In addition, the production of secondary neutrons was measured and estimated as a function of the distance between the isocenter and detector. The neutron dose was measured using WENDI-II (Wide Energy Neutron Detection Instruments) and a Plastic Water phantom; a Zebra dosimeter and 4-cm-thick range shifter were also employed as a phantom. In conclusion, we need to consider the secondary neutron dose at proton scanning facilities to employ the range shifter reasonably and effectively.

• Journal of Radiation Protection and Research
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• v.42 no.3
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• pp.141-145
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• 2017

Background: The concrete walls inside the vaults of cyclotron facilities are activated by neutrons emitted by the targets during radioisotope production. Reducing the amount of radioactive waste created in such facilities is very important in case they are decommissioned. Thus, we proposed a strategy of reducing the neutron activation of the concrete walls in cyclotrons during operation. Materials and Methods: A polyethylene plate and B-doped Al sheet (30 wt% of B and 2.5 mm in thickness) were placed in front of the wall in the cyclotron room of a radioisotope production facility for pharmaceutical use. The target was Xe gas, and a Cu block was utilized for proton dumping. The irradiation time, proton energy, and beam current were 8 hours, 30 MeV, and $125{\mu}A$, respectively. To determine a suitable thickness for the polyethylene plate set in front of the B-doped Al sheet, the neutron-reducing effects achieved by inserting such sheets at several depths within polyethylene plate stacks were evaluated. The neutron fluence was monitored using an activation detector and 20-g on de Au foil samples with and without 0.5-mm-thick Cd foil. Each Au foil sample was pasted onto the center of a polyethylene plate and B-doped Al sheet, and the absolute activity of one Au foil sample was measured as a standard using a Ge detector. The resulting relative activities were obtained by calculating the ratio of the photostimulated luminescence of each foil sample to that of the standard Au foil. Results and Discussion: When the combination of a 4-cm-thick polyethylene plate and B-doped Al sheet was employed, the thermal neutron rate was reduced by 78%. Conclusion: The combination of a 4-cm-thick polyethylene plate and B-doped Al sheet effectively reduced the neutron activation of the investigated concrete wall.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.745-752
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• 2024

A method for gamma/neutron event classification based on frequency-domain analysis for mixed radiation environments is proposed. In contrast to the traditional charge comparison method for pulse-shape discrimination, which requires baseline removal and pulse alignment, our method does not need any preprocessing of the digitized data, apart from removing saturated traces in sporadic pile-up scenarios. It also features the identification of neutron events in the detector's full energy range with a single device, from thermal neutrons to fast neutrons, including low-energy pulses, and still provides a superior figure-of-merit for classification. The proposed frequency-domain analysis consists of computing the fast Fourier transform of a triggered trace and integrating it through a simplified version of the transform magnitude components that distinguish the neutron features from those of the gamma photons. Owing to this simplification, the proposed method may be easily ported to a real-time embedded deployment based on Field-Programmable Gate Arrays or Digital Signal Processors. We target an off-the-shelf detector based on a small CLYC (Cs2LiYCl6:Ce) crystal coupled to a silicon photomultiplier with an integrated bias and preamplifier, aiming at lightweight embedded mixed radiation monitors and dosimeter applications.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.155-163
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• 2020

To investigate the optimal detector material for prompt gamma imaging during boron neutron capture therapy, in this study, we evaluated the characteristic regarding radiation reaction of available detector materials using a Monte Carlo simulation. Sixteen detector materials used for radiation detection were investigated to assess their advantages and drawbacks. The estimations used previous experimental data to build the simulation codes. The energy resolution and detection efficiency of each material was investigated, and prompt gamma images during BNCT simulation were acquired using only the detectors that showed good performance in our preliminary data. From the simulation, we could evaluate the majority of detector materials in BNCT and also could acquire a prompt gamma image using the six high ranked-detector materials and lutetium yttrium oxyorthosilicate. We provide a strategy to select an optimal detector material for the prompt gamma imaging during BNCT with three conclusions.

• Bulletin of the Korean Chemical Society
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• v.27 no.10
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• pp.1609-1612
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• 2006

The determination of the boron isotopic ratio in solutions was achieved by means of a solid state track detector by using an alpha track. The neutron flux was optimized by using a Cd-foil to find the optimum conditions for counting the number of alpha tracks on the selected solid detector caused by the (n, $\alpha$) nuclear reaction of boron. The home-made multi-dot detector plate was utilized in this study to increase the reproducibility of the measurement by uniformly drying the boron solution within the marked circle area on the detector plate. The experimental results of this study verified that the $^{11}B/^{10}B $ isotopic ratio can be measured by observing the number of alpha tracks for different concentrated standard solutions with various isotopic compositions. This technique was applied to the determination of $^{10}B$ enrichment factor in a biological sample for a boron neutron capture therapy.

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1942-1946
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• 2021

Gadolinium oxysulfide (GOS) is regarded as a novel scintillator for the realization of ultra-high spatial resolution in neutron imaging. Monte Carlo simulations of GOS scintillator show that the capability of its spatial resolution is towards the micron level. Through the time-of-flight method, the light output of a GOS scintillator was measured to be 217 photons per captured neutron, ~100 times lower than that of a ZnS/LiF:Ag scintillator. A detector prototype has been developed to evaluate the imaging solution with the GOS scintillator by neutron beam tests. The measured spatial resolution is ~36 ㎛ (28 line pairs/mm) at the modulation transfer function (MTF) of 10%, mainly limited by the low experimental collimation ratio of the beamline. The weak light output of the GOS scintillator requires an enormous increase in the neutron flux to reduce the exposure time for practical applications.

• Journal of Radiation Protection and Research
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• v.26 no.3
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• pp.249-253
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• 2001

Neutron dose assessment in criticality accidents using Whole Body Counter (WBC) was proved to be an effective method as rapid neutron dose estimation at the JCO criticality accident in Tokai-mura. The 1.36MeV gamma-ray of $^{24}Na$ in a body can be detected easily by a germanium detector. The Minimum Detectable Activity (MDA) of $^{24}Na$ is approximately 50Bq for 10miniute measurement by the germanium-type whole body counter at JNC Tokai Works. Neutron energy spectra at the typical shielding conditions in criticality accidents were calculated and the conversion factor, whole body activity-to-organ mass weighted neutron absorbed dose, corresponding to each condition were determined. The conversion factor for uncollied fission spectrum is 7.7 $[(Bq^{24}Na/g^{23}Na)/mGy]$.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.2984-2996
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• 2023

A 100 kW thermal power pool-type light water reactor and Pu(Be) as a fast neutron source were used to determine the appropriate carrier for irradiating boron-containing samples with neutron beams. The tested materials (carriers) were subjected to neutron beams in the reactor's tangential channel. The geometrical arrangement of experimental facilities relative to the neutron beam trajectory, as well as the effect of sample thickness on the count rate, were investigated. The majority of the detectable charged particles emitted by the neutron beam's interaction with tested materials and the detector's detecting layer are protons (recoiled hydrogen) and particles generated in nuclear reactions (protons and alpha particles), respectively. Stopping and Range of Ions in Matter (SRIM) software was used to do theoretical calculations for the range of expected released particles in various materials, including human tissue. The results of measurement and calculation are in good agreement. According to experiments and theoretical calculations, the number of protons emitted by tissue-like materials may commit a dose comparable to that of boron capture reactions. Furthermore, the range of protons is significantly larger than that of alpha particles, which most probably changes dose distribution in healthy cells surrounding the tumor, which is undesirable in the BNCT approach.