• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.801-809
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• 2017

The purpose of this study is to assess the additional neutron effective dose during passive scattering proton therapy. Monte Carlo code (Monte Carlo N-Particle 6) simulation was conducted based on a precise modeling of the National Cancer Center's proton therapy facility. A three-dimensional neutron effective dose profile of the interior of the treatment room was acquired via a computer simulation of the 217.8-MeV proton beam. Measurements were taken with a $^3He$ neutron detector to support the simulation results, which were lower than the simulation results by 16% on average. The secondary photon dose was about 0.8% of the neutron dose. The dominant neutron source was deduced based on flux calculation. The secondary neutron effective dose per proton absorbed dose ranged from $4.942{\pm}0.031mSv/Gy$ at the end of the field to $0.324{\pm}0.006mSv/Gy$ at 150 cm in axial distance.

• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1406-1413
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• 2022

In this study, a diamond-based neutron scatter camera (DNSC) was developed for neutron spectroscopy in high flux environments. The DNSC was evaluated experimentally and through simulations. It was simulated using several Monte Carlo codes in a two-array layout. The two-array model included two diamond detectors. The simulation reconstructed the spectra of ²⁵²Cf and ²³⁹Pu-Be neutron sources with high accuracy (~93%). The two-diamond array system was experimentally evaluated, demonstrating the neutron spectroscopy capabilities of the DNSC. The reconstructed spectrum of the ²³⁹Pu-Be source manifested the characteristic peaks of the source. The advantage of a DNSC over a NSC is its ability to define any neutron double-scattering events without the need to absorb incident neutrons in the second detector, and atomic recoil energy information is not needed to determine the incident neutron energy.

• Journal of Radiation Protection and Research
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• v.37 no.2
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• pp.70-74
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• 2012

Silicon carbide (SiC) is a promising material for neutron detection at harsh environments because of its capability to withstand strong radiation fields and high temperatures. Two PIN-type SiC semiconductor neutron detectors, which can be used for nuclear power plant (NPP) applications, such as in-core reactor neutron flux monitoring and measurement, were designed and fabricated. As a preliminary test, MCNPX simulations were performed to estimate reaction probabilities with respect to neutron energies. In the experiment, I-V curves were measured to confirm the diode characteristic of the detectors, and pulse height spectra were measured for neutron responses by using a $^{252}Cf$ neutron source at KRISS (Korea Research Institute of Standards and Science), and a Tandem accelerator at KIGAM (Korea Institute of Geoscience and Mineral Resources). The neutron counts of the detector were linearly increased as the incident neutron flux got larger.

• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.277-284
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• 2017

This paper presents the establishment and characterization of a neutron calibration field using a bare $^{252}Cf$ source of low neutron source strength in Vietnam. The characterization of the field in terms of neutron flux spectra and neutron ambient dose equivalent rates were performed by Monte Carlo simulations using the MCNP5 code. The anisotropy effect of the source was also investigated. The neutron ambient dose equivalent rates at three reference distances of 75, 125, and 150 cm from the source were calculated and compared with the measurements using the Aloka TPS-451C neutron survey meters. The discrepancy between the calculated and measured values is found to be about 10%. To separate the scattered and the direct components from the total neutron flux spectra, an in-house shadow cone of 10% borated polyethylene was used. The shielding efficiency of the shadow cone was estimated using the MCNP5 code. The results confirmed that the shielding efficiency of the shadow cone is acceptable.

• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1479-1486
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• 2019

The purpose of the present study is to develop the 3D static and noise simulator based on Galerkin Finite Element Method (GFEM) using the unstructured hexahedral elements. The 3D, 2G neutron diffusion and noise equations are discretized using the unstructured hexahedral by considering the linear approximation of the shape function in each element. The validation of the static calculation is performed via comparison between calculated results and reported data for the VVER-1000 benchmark problem. A sensitivity analysis of the calculation to the element type (unstructured hexahedral or tetrahedron elements) is done. Finally, the neutron noise calculation is performed for the neutron noise source of type of variable strength using the Green function technique. It is shown that the error reduction in the static calculation is considerable when the unstructured tetrahedron elements are replaced with the hexahedral ones. Since the neutron flux distribution and neutron multiplication factor are appeared in the neutron noise equation, the more accurate calculation of these parameters leads to obtaining the neutron noise distribution with high accuracy. The investigation of the changes of the neutron noise distribution in axial direction of the reactor core shows that the 3D neutron noise analysis is required instead of 2D.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4226-4230
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• 2022

The Korea Atomic Energy Research Institute (KAERI) has designed a Hybrid-Low Power Research Reactor (H-LPRR) which can be used for critical assembly and conventional research reactor as well. It is an open tank-in-pool type research reactor (Thermal Power: 50 kW_th) of which the most important applications are Neutron Activation Analysis (NAA), Radioisotope (RI) production, education and training. There are eight irradiation holes on the edge of the reactor core: IR (6 holes for RI production) and NA (2 holes for NAA) holes. In order to quantify the elemental concentration in target samples through the Instrumental Neutron Activation Analysis (INAA), it is necessary to measure neutron spectrum parameters such as thermal neutron flux, the deviation from the ideal 1/E epithermal neutron flux distribution (α), and the thermal-to-epithermal neutron flux ratio (f) for the irradiation holes. In this study, the MCNP6.1 code and FORTRAN 90 language are applied to determine the parameters for the two irradiation holes (NA-SW and NA-NW) in H-LPRR, and in particular its α and f parameters are compared to values of other research reactors. The results confirmed that the neutron irradiation holes in H-LPRR are designed to be sufficiently applied to neutron activation analysis, and its performance is comparable to that of foreign research reactors including the TRIGA MARK II.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2419-2431
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• 2023

Experiments related to Boron Neutron Capture Therapy (BNCT) accomplished at the Institute of Nuclear Techniques (INT), Budapest University of Technology and Economics (TUB) are presented. Relevant investigations are required before designing BNCT for vivo applications. Samples of relevant boron compounds (H₃BO₃, BDTPA) usually employed in BNCT were investigated with neutron beam. Channel #5 in the research reactor (100 kW) of INT-TUB provides the neutron beam. Boron samples are mounted on a carrier for neutron irradiation. The particle attenuation of several carrier materials was investigated, and the one with the lowest attenuation was selected. The effects of boron compound type, mass, and compound phase state were also investigated. To detect the emitted charged particles, a traditional ZnS(Ag) detector was employed. The neutron beam's interaction with the detector-detecting layer is investigated. Graphite (as a moderator) was employed to change the neutron beam's characteristics. The fast neutron beam was also thermalized by placing a portable fast neutron source in a paraffin container and irradiating the H₃BO₃. The obtained results suggest that the direct measurement approach appears to be insufficiently sensitive for determining the radiation dose committed by the Alpha particles from the ¹⁰B (n,α) reaction. As a result, a new approach must be used.

• Journal of Radiation Protection and Research
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• v.41 no.4
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• pp.384-388
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• 2016

Background: For a verification of newly-developed neutron absorbers, one of guidelines on the qualification and acceptance of neutron absorbers is the neutron attenuation test. However, this approach can cause a problem for the qualifications that it cannot distinguish how the neutron attenuates from materials. Materials and Methods: In this study, an estimation method of neutron absorption performances for materials is proposed to detect both direct penetration and back-scattering neutrons. For the verification of the proposed method, MCNP simulations with the experimental system designed in this study were pursued using the polyethylene, iron, normal glass and the vitrified form. Results and Discussion: The results show that it can easily test neutron absorption ability using single absorber model. Also, from simulation results of single absorber and double absorbers model, it is verified that the proposed method can evaluate not only the direct thermal neutrons passing through materials, but also the scattered neutrons reflected to the materials. Therefore, the neutron absorption performances can be accurately estimated using the proposed method comparing with the conventional neutron attenuation test. Conclusion: It is expected that the proposed method can contribute to increase the reliability of the performance of neutron absorbers.

• Progress in Superconductivity and Cryogenics
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• v.16 no.1
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• pp.9-13
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• 2014

Effects of neutron irradiation on the superconducting properties of the undoped $MgB_2$ and the carbon(C)-doped $MgB_2$ bulk superconductors, prepared by an in situ reaction process using Mg and B powder, were investigated. The prepared $MgB_2$ samples were neutron-irradiated at the neutron fluence of $10^{16}-10^{18}n/cm^2$ in a Hanaro nuclear reactor of KAERI involving both fast and thermal neutron. The magnetic moment-temperature (M-T) and magnetization-magnetic field (M-H) curves before/after irradiation were obtained using magnetic property measurement system (MPMS). The superconducting critical temperature ($T_c$) and transition width were estimated from the M-T curves and critical current density ($J_c$) was estimated from the M-H curves using a Bean's critical model. The $T_cs$ of the undoped $MgB_2$ and C-doped $MgB_2$ before irradiation were 36.9-37.0 K and 36.6-36.8 K, respectively. The $T_cs$ decreased to 33.2 K and 31.6 K, respectively after irradiation at neutron fluence of $7.16{\times}10^{17}n/cm^2$, and decreased to 22.6 K and 24.0 K, respectively, at $3.13{\times}10^{18}n/cm^2$. The $J_c$ cross-over was observed at the high magnetic field of 5.2 T for the undoped $MgB_2$ irradiated at $7.16{\times}10^{17}n/cm^2$. The $T_c$ and $J_c$ variation after the neutron irradiation at various neutron fluences were explained in terms of the defect formation in the superconducting matrix by neutron irradiation.

• 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.