• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 추계학술발표회요약집
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• pp.387.1-387
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• 2002

• Nuclear Engineering and Technology
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• 제29권2호
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• pp.148-157
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• 1997

A new capsule that has a unique structure in which the test environments including neutron flux and fluence, and irradiation temperature can be controlled precisely during irradiation, was conceptually designed. The capsule structure and instrumentation were successfully designed according to the JMTR's standard procedures of capsule design. Based on the target irradiation, the details of the irradiation such as neutron fluence and irradiation temperature ore calculated and the related capsule safety was evaluated. In addition, the effects of design parameters including the changes in inner-capsule configuration, heater capacity, and Helium gas pressure on the specimen temperature were analyzed with a computer program. Through these thermal and strength evaluations, this capsule was proved to be safe during the irradiation in the JMTR.

• Journal of Radiation Protection and Research
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• 제26권3호
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• pp.315-319
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• 2001

It is shown, that both cosmic radiation (external source) and natural radioactive gases (inner source) are sources of neutrons near the Earth crust. Correlation between the Earth crust dynamics and variations of thermal and slow neutron flux near the Earth surface is studied. It is shown, that variations of neutron flux near the Earth crust can be used for short-term predicting of natural hazards.

• Nuclear Engineering and Technology
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• 제45권2호
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• pp.219-222
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• 2013

Alloy 690 has been selected as a steam generator tubing material for SMART owing to a near immunity to primary water stress corrosion cracking. The steam generators of SMART are faced with a neutron flux due to the integrated arrangement inside a reactor vessel, and thus it is important to know the irradiation effects of the thermal conductivity of Alloy 690. Alloy 690 was irradiated at HANARO to fluences of (0.7-28) ${\times}10^{19}n/cm^2$ (E>0.1MeV) at $250^{\circ}C$, and its thermal conductivity was measured using the laser-flash equipment in the IMEF. The thermal conductivity of Alloy 690 was dependent on temperature, and it was a good fit to the Smith-Palmer equation, which modified the Wiedemann-Franz law. The irradiation at $250^{\circ}C$ did not degrade the thermal conductivity of Alloy 690, and even showed a small increase (1%) at fluences of (0.7~28) ${\times}10^{19}n/cm^2$ (E>0.1MeV).

• Nuclear Engineering and Technology
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• 제49권4호
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• pp.781-786
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• 2017

The presence of crust on the inner walls of metallic ducts impairs transportation because crust completely or partially hinders the passage of fluid to the processing unit and causes damage to equipment connected to the production line. Its localization is crucial. With the development of the electronic imaging system installed at the Argonauta/Nuclear Engineering Institute (IEN)/National Nuclear Energy Commission (CNEN) reactor, it became possible to visualize crust in the interior of metallic piping of small diameter using real-time neutron radiography images obtained with a low neutron flux. The obtained images showed the resistance offered by crust on the passage of water inside the pipe. No discrepancy of the flow profile at the bottom of the pipe, before the crust region, was registered. However, after the passage of liquid through the pipe, images of the disturbances of the flow were clear and discrepancies in the flow profile were steep. This shows that this technique added the assembled apparatus was efficient for the visualization of the crust and of the two-phase flows.

• Nuclear Engineering and Technology
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• 제55권11호
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• pp.4246-4251
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• 2023

The optimized design of a Neutron Activation Analysis (NAA) system, including Delayed Gamma NAA (DGNAA) and Prompt Gamma NAA (PGNAA), has been proposed in this research based on Mevex Linac with 5 MeV electron energy and 50 kW power as a neutron source. Based on the MCNPX 2.6 simulation, the optimized configuration contains; tungsten as an electron-photon converter, BeO as a photoneutron target, BeD₂ and plexiglass as moderators, and graphite as a reflector and collimator, as well as lead as a gamma shield. The obtained thermal neutron flux at the beam port is equal to 2.06 × 10⁹ (# /cm².s). In addition, using the optimized neutron beam, the detection limit has been calculated for some elements such as H-1, B-10, Na-23, Al-27, and Ti-48. The HPGe Coaxial detector has been used to measure gamma rays emitted by nuclides in the sample. By the results, the proposed system can be an appropriate solution to measure the concentration and toxicity of elements in different samples such as food, soil, and plant samples.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4329-4334
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• 2023

In this research, a prompt gamma neutron activation analysis (PGNAA) system is designed and constructed based on the use of a low power research reactor. For this purpose, despite the fact that this reactor did not include beam tubes, a thermal neutron beam line is installed inside the reactor tank. The extraction of the beam line from inside the tank made it possible to provide the neutron flux from the order of 106 n.cm^-2.s^-1. Also, because the beam line is installed in a tangential position to the reactor core, its gamma level has been minimized. Also, a suitable radiation shield is considered for the detector to minimize the background radiation and prevent radiation damage to the detector. Calculations and measurements are done in order to characterize this system, as well as spectrometry of several samples. The results of evaluations and experiments show that this system is suitable for performing PGNAA.

• Nuclear Engineering and Technology
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• 제53권6호
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• pp.1970-1978
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• 2021

The aim of this study was to optimize the target, moderator, and collimator (TMC) in a neutron beam generator for the accelerator-based BNCT (A-BNCT) system. The optimization employed the Monte Carlo Neutron and Photon (MCNP) simulation. The optimal geometry for the target was decided as the one with the highest neutron flux among nominates, which were called as angled, rib, and tube in this study. The moderator was optimized in terms of consisting material to produce appropriate neutron energy distribution for the treatment. The optimization of the collimator, which wrapped around the target, was carried out by deciding the material to effectively prevent the leakage radiations. As results, characteristic of the neutron beam from the optimized TMC was compared to the recommendation by the International Atomic Energy Agent (IAEA). The tube type target showed the highest neutron flux among nominates. The optimal material for the moderator and collimator were combination of Fluental (Al₂0₃+AlF₃) with ⁶⁰Ni filter and lead, respectively. The optimized TMC satisfied the IAEA recommendations such as the minimum production rate of epithermal neutrons from thermal neutrons: that was 2.5 times higher. The results can be used as source terms for shielding designs of treatment rooms.

• Nuclear Engineering and Technology
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• 제38권6호
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• pp.585-590
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• 2006

This paper describes the results of an irradiation test and the specifications of the pneumatic transfer system (PTS) in the NAA #3 irradiation hole at the HANARO research reactor, which was reinstalled after some modifications of the operation mode at the end of 2004. The outer and inner diameters of the PE transfer tube are 34.1 and 27.5 mm, respectively. PE rabbit was used for sample irradiation. The $N_2$ gas pressure of the PTS lines was adjusted to 0.75 bar. The average sending time to the reactor was $8.5{\pm}0.3$ s and the average receiving time back to the receiver was $3.2{\pm}0.2$ s. The internal and external temperature of the irradiation tube was measured in a range of 50 to $80^{\circ}C$ for a 40 s to 80 s irradiation time, respectively. The optimum irradiation time was estimated to be less than 80 s. The thermal, epithermal and fast neutron flux at 30 MW thermal power were $1.42{\pm}0.01{\times}10^{14},\;1.51{\pm}0.04{\times}10^{13}$ and $9.48{\pm}0.69{\times}10^{11} n{\cdot}cm^{-2}{\codt}s^{1-}$, respectively. The cadmium ratio was approximately 9.40. The data obtained will be applied to supplement user information and for reactor management.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3413-3420
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• 2021

Applying the available neutron flux for medical and industrial purposes is the most important application of research reactors. The neutron radiography system is used for non-destructive testing (NDT) of materials so that it is one of the main applications of nuclear research reactors. One of these research reactors is the 5 MW pool-type light water research reactor of Tehran (TRR). This work aims to investigate on materials and location of the beam tube (BT) of the TRR radiography system to improve the index parameters of BT. Our results showed that a through-type BT with 20 cm thick carbon neutron filter, 1.2 cm and 9.4 cm of the diameter of inlet (D₁) and output (D₂) BT, respectively gives thermal neutron flux almost 25.7, 5.6 and 1.1 times greater than the former design of the TRR (with D₁ = 1.8 cm and D₁ = 9.4 cm), previous design of the TRR with D₁ = 3 cm and D₁ = 9.4 cm, and another design with D₁ = 5 cm and D₁ = 9.4 cm, respectively. Therefore, the design proposed in this paper could be a better alternative to the current BT of the TRR.